VN9700001

VIEN NANG LU0NG NGUYEN TLf VIET NAM

CAC BAC TRL/NG VAT LY LO• •

CUA LO PHAN UfNG HAT NHAN DA LAT

TUYEN TAP CAC BAO CAO KHOA HQC VA CAC CONG TRINH KHOA HQCNHAN DIP KY NI$M Mlfdl NAM KHANH THANH LO PHAN (fNG HAT NHAN DA LA T

20/3/1984 - 20/3/1994

Bien soanPGS, PTS NGO QUANG HUY

Trung tarn ky thuat hat nhan TP Ho Chi Minh

TP Ho Chi Minh-i0/1994

28 t

POOR QUALITY]0RICHNA1.

We regret thatsome of the pagesin this report may

not be up to theproper legibilitystandards, eventhough the best

possible copy wasused for scanning

BAU

L6 phan ufng hat nhan Da lat (LPUHNDL) dUdc khoi phyc tu1 16 TRIGAMark II, bat d'au hoat dpng nam 1963 vdi cong suat danh djnh 250 kW. Nam1975, khi rut quan kh6i mien Nam, My da mang di net cac thanh nhien lieuuranium lam cho 16 phan ufng (LPU) khong hoat d6ng du"dc nQa.

DUdc sU giup dcf cua Lien x6 (cu), cdng trlnh khoi phuc va mo" rdngLPUHNDL dUdc tien hanh trong cac nam 1982 -1984. Ngay 1/11/1983LPUHNDL ducfc nap nhien lieu d£n trang thai tcfi nan. Sau mot thdi gian kh6idong vat ly va khdi dpng nang lUpfng, ngay 20/3/1984 LPli da chinh thufc duravao hoat d6ng vdi cong suat thiet k§' 500 kW.

Viec khoi phuc LPUHNDL thi/c chat la trang b| lai hau het cac he c6ngnghe. Cac thanh ph'an cu cua 16 TRIGA Mark II cdn giuf lai la vanh phan xagraphite, thung 16 b&ng nhom, 4 kdnh ngang thf nghiem, c6t nhiet va tu"dngbe tong can xa sinh hoc. Cac he thd'ng cong nghe mdi cua Lien x'6 dap ufngnhu c'au nang cong suat 16 len gap doi, 500 kW, va nang cao mufc dp antoan bufc xa. Cac he nay g'6m vung hoat, h§ thong tai nhiet cho nude LPU,he thong dieu khien va bao v§, he thong kiem tra an toan bijfc xa, he thongthai khi phong xa, he thong xu" ly chat thai phong xa I6ng, he thong ch6n catcac ba thai phong xa, he thong cung cap dien cong nghiep Uu tien va diensu" co ...

Trong cac he cong ngh^ neu trdn, vung hoat cua LPU la thanh phanquan trong nhat. Vung hoat mdi su" dung cac thanh nhien lieu loai VVR-M2cua Lien xo, la hdp kirn uranium-nhom, dat Ipt vao trong vanh phan xagraphite cu cua 16 TRIGA Mark II. M6t lu"dng beryllium diidc dat vao tarnvung hoat, tao nen mot bay n-dtr&n, va d m^p ngoai vung hoat, tao nen motvanh phan xa bo sung.

LPUHNDL la LPU duy nhat tren the gidi lai tao giufa loai 16 TRIGA cuaMy va loai 16 VVR cua Lien xo. Dufng ve phUdng dien vat ly 16 phan ufng,LPUHNDL co ba dac diem chfnh sau day khac vdi 16 TRIGA:

1. Cac thanh nhien lieu loai U-ZrHn cua 16 TRIGA, co he so phan hoiam lefn theo nhiet dp, dUdc thay b^ng cac thanh nhien lieu VVR-M2, la hdpkirn uranium-nhom co he so phan hoi am be theo nhiet dp. Dieu do co nghTa

la thay vung hoat co tfnh chat an toan npi tai cao b^ng vung hoat co tfnh antoan npi tai thap hdn.

2. Viec dtfa vao vung hoat mot bay ncftron 6" tarn va mot vanh phan xaberyllium bo sung 6 vanh ngoai lam nang cao m&t dp thong lu"0ng ndtron 6'tarn vung hoat va lam cho phan bo tracing ndtron trong vung hoat trd nenphufc tap hdn.

3. Viec dJa them vat lieu beryllium vao vung hoat dan den si/ xuathien cac photondtron tre do tUdng tac cua c&c gamma n&ng liiOng tren 1,66MeV v6i v&t lieu nay. Cac photondtron tre co the c6 anh hu6ng den dpnghpc LPU va v&n hanh LPU.

Viec nghien cufu cac dac diem nay cua LPL/HNDL la van 6e ly thu catu' phudng dien vat ly LPU Ian phadng didn bao dam van hanh an toan 16.Trong tuye'n tap nay tap hdp mot so bao c&o khoa hpc va cong trinh khoahpc cua t^p the cac can bp khoa hpc Vi#n Nghi§n cufu hat nhan Da lat vaTrung tarn ky thuat hat nhan TP Ho Chf Minh th^c hien trong 10 n^m 1984 -1994. Npi dung tuyen tap gbm hai phan chfnh:

1. Bao cao khdi dpng LPliHNDL, trinh bay cac ket qua khdi dpng LPUtCr thang 11/1983 den thang 2/1984. Bao cao gbm 3 phan:

a. Khdi dpng vat ly vdi cau hlnh vung hoat khong c6 bay ndtron, nhlmxac dinh-khoi Udng tdi han cua vung hoat.

b. Khdi dpng vat ly vdi cau hlnh vung hoat co bay ndtron, nh&m taovung hoat lam viec sau nay va x&c dinh cac dac tri/ng vat ly.

c. Kh6i dpng nang lUdng, xac dinh va nang cong suat 16, van hanh 6cong suat danh djnh va x&c dinh cac dac triing nhiet thuy dpng.

2. Cac cong trinh nghien cufu ve cac thong so vat ly 16 va nhiet kythuat trong qua trinh hoat dpng cua 16 trong 10 n&m qua. C^c ket qua thi/cnghiem va tinh toan ly thuyet gbm : xac dinh phan hieu dung va thanh ph'ancac photondtron tre, ngubn ph6tondtr6n tre con lai trong 16 sau khi dC/ng 16,cac mufc cong suat tdi han toi thieu cua 16, hieu ufng nhieu loan tracing ndtrontrong vung hoat khi thay thanh nhien lieu being vat lieu khac, cac phan botheo chieu cao va theo ban kfnh vung hoat cua ndtron nhiet, d6 phan ufngdtfdng CL/C dai du"a vao vung hoat ma khong gay SL/ co ve cong suat, he so

nhet dp cua dp phan ufng doi vdi nUdc, hieu ufng nhiem doc Xe non, xac dinhnhiet dp be mat cac thanh nhien lieu, cac dac trUng thuy nhiet trong cac dieukien chuyen tiep, tinh toan dp chay cua nhien lieu ...

Cac bao cao va cong trlnh khoa hoc trong tuyen tap nay, la mot phantrong nhieu ket qua nghien cufu vat ly 16 phan ufng doi vdi LPUHNDL cuaVien Nghien cufu hat nhan Da lat va Trung tarn Ky thuat hat nhan TPHCMtrong 10 nam qua, nham lam sang t6 mot vai khia canh ve ba tinh chat dacthu cua LPUHNDL nha da neu tren. Doc gia co the tham khao cac cong trlnhkhac dUdc cong bo trong nude va ngoai nude de cd mot cai nhin tong quatva day du hdn. Tuy nhien viec tap hdp tiidng doi cd he thong mot so bao caova cong trinh nay hy vpng ddng gdp di/dc mot phan nh6 de gidi thieu cachoat dpng nhi'eu mat cua LPU trong 10 nSm qua va danh dau ngay ky niem10 nam khanh thanh LPUHNDL.

De cd dUdc cac bao cao khoa hoc va cac cong trinh khoa hoc trongcac nam qua tren LPUHNDL, chung toi xin cam dn Ban Giam doc VienNang lUdng nguyen tCr quoc gia ( nay la Vien NSng lUdng nguyen tC/ Vietnam) va Ban Giam doc Vien Nghien cufu hat nhan Da lat trong vi§c lanh daothanh cong khoi phuc va diJa LPU vao hoat dpng. Vdi tinh cam men phuc,chung t6i xin cam dn cac anh chj da tham gia khoi phuc LPU trong dieukien khd khan cua nhufng nSm dau th&p ky 80. Xin cam dn Ph6ng 16 phanufng va cac kfp van hanh 16 tao nhi'eu dieu kien thuan Idi va giup d5 tienhanh thi nghiem, Phong Vat ly hat nhan ufng dung va Phong an toan bufc xatrong viec cho phep su" dung cdc he do bufc xa. Chung t6i xin chan thanhcam dn cac nha khoa hpc va ky thuat da thao luan va cd nhieu y kien s^usSc trong nhi'§u nam qua dd'i vdi npi dung cac cong trlnh neu trong tuyen tapva mong nhan dUdc nhufng y ki^n ddng gdp tiep theo trong thdi gian tdi vituyen tap nay chSc ch in con nhieu sai sot.

BAO CAOKHdl DONG LO PHAN LfNG HAT NHAN DA LAT

Pham Duy Hien, Ngo Quang Huy, Vu Ha*i Long, TrSn Khanh Mai

N0I DUNGTrang

BAO CAO KHCJl B0NG Ld PHAN QNG HAT NHAN DA LATPham Duy Hien, Ngo Quang Huy, VQ Hai Long, Tran Khanh Mai 1'4

Phan 1: K77O7 dong vat ly Id phan ufng hat nhan Da lat vdi cau hinhvung hoat khong co bay natron ^-22

Phan 2: Khdi dong vat ly 16 phan ufng hat nhan Da lat vdi cau hinhvung hoat co bay natron

Phan 3: Khdi dong nang IU0ng Id phan dng hat nhan Da lat O. . n oO I ~ I \JC.

CAC CONG TRJNH KHOA HQC NGHIEN CtfU CAC DJ\C TRUNG 103-104VAT LY LO CUA LO PHAN UfNG HAT NHAN DA LAT

Preliminary experiments performed on the newly reconstructed ^ Qg.-j ^ %500 kW nuclear research reactor of DalatPham Duy Hien and Ngo Quang Huy

Neutron spectra of the newly reconstructed 500 kW nuclear reactor of 113-120Dalat, Vietnam. Application in activation analysisPham Duy Hien.Tran Khanh Mai, Ngo Quang Huy, Nguyen Tac Anh

Investigation of the basic reactor physics characteristics of the Dalat 121-128nuclear research reactorNgo Quang Huy, Ha Van Thong and Ngo Phu Khang

Delayed photoneutrons of the Dalat nuclear research reactor • 129-136Ngo Quang Huy, Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Nhi Dien, Pham Van Lam, Huynh Dong Phuong,Luong Ba Vien, Le Vinh Vinh

Determination of the lowest critical power levels of the Dalat nuclearresearch reactor 137-144Ngo Quang Huy, Ha Van Thong, Vu Hai Long, Do Quang Binh, HuynhTon Nghiem, Nguyen Minh Tuan, Luong Ba Vien, Le Vinh Vinh

The neutron field perturbation effect in the Dalat reactorNgo Quang Huy, Ha Van Thong, Vu Hai Long, Ngo Phu Khang, 145-154Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh

Trang

Effective height of the core of the Dalat nuclear research reactor 155-162Ngo Quang Huy, Ha Van Thong, Vu Hai Long, Nguyen Due Binh,Nguyen Minh Tuan, Luong Ba Vien, Le Vinh Vinh, David Perez Martin,Fernando Garcia Yip

The radial distribution of the neutron field in the core of the Dalat reactor ^ 63-168Ngo Quang Huy, Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh

Distribution of the thermal neutron field around the graphite reflector of1 69-176

the Dalat nuclear research reactorNgo Quang Huy, Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh

Calculation of fuel burn-up and fuel reloading for the Dalat nuclearresearch reactorNguyen Phuoc Lan.Ngo Quang Huy,Ha Van Thong,Do Quang Binh

Kinetic characteristics of the Dalat nuclear research reactor . g g . ggTran Khac An, Nguyen Nhi Dien, Pham Duy Hien, Ngo Quang Huy,Ngo Phu Khang, Pham Van Lam, Vu Hai Long, Huynh Dong Phuong,Ha Van Thong, Luong Ba Vien, Le Vinh Vinh

Experimental determination of fuel surface temperatures in the Dalat ^ 97-206nuclear research reactorNgo Phu Khang, Ngo Quang Huy, Tran Khac An, Pham Van Lam,P.A. Diakov, A.F. Sacsanov

Thermohydraulic characteristics under some transient conditions of the 207-218Dalat nuclear research reactorNgo Quang Huy, Ngo Phu Khang, Tran Khac An, Huynh Ton Nghiem

Experimental investigation of the neutron physics characteristics of the 219-231Dalat nuclear research reactorNgo Quang Huy and Ha Van Thong

BAO CAOKHdl DONG LO PHAN LTNG HAT NHAN DA LAT

Trong B&o cao nay trlnh bay ket qua khdi dpng 16 phan ufng hatnhan Da I at, gorn ba ph&n:

- Phan 1: Khdi ddng v&t ly vdi cau hlnh vung hoat khong co bayndtron, xac djnh khoi lifdng tdi nan.

- Phan 2: Khdi dpng vat ly vdi cau hlnh vung hoat c6 bay ncftron,xac dinh mot so dac trUng vat ly.

- Phan 3: Khdi dpng na"ng lu"0ng, van hanh d cong suat danh dinh,xac dinh dac trifng nhiet thuy d6ng.

Sau thdi gian khdi d6ng L6 phan ufng, nh6m khdi dpng da bi§n soan(bing tieng Nga) cac tai li#u nghi#m thu ky thuat va bi^n ban t6m t&t ca"c ketqua vat ly. Ban bao c&o nay t6ng hp*p c^c du lieu ghi ch^p trong qu& trlnhkhdi dpng, dac biet trong Nhat ky khdi d6ng va Van hanh L6 phan ufng.

CAC CAN BQ THAM GIA TRl/C TIEP KHdl DQNG LO PHAN JNG

/. Phfa Viet nam :

Pham Duy Hien, Ng6 Quang Huy, Vu Hai Long, Tr'an Khanh Mai,Nguyen B£ng NhuSn, Tr^n KhSc An, Nguyen Nhi Di'dn, Nguy&n Ngpc Lam,Do van Hiep, Pham Xuan Phacfng, Nguyen TrUdng Sinh, Pham v^n Lam,Dudng Quang Tan, Tradng Cam Ranh, Nguyen T^c Anh, Ladng NgocChau, Le VTnh Vinh, Huynh D6ng PhUdng, Nguyen Th| N£ng, TO vSn NghTa.

2. Phfa Lien xo (cu):

Arhanghenski N.V., Terekhov A.V., Kuznexov E.M., Trophimov I.V.,Lavrenov N.P., Kharlamov N.B., Sokolov I.V., Bakharev V.S., Truskin V.S.,Khandamirov U.E., Semenov S.M.

CAC CAN BQ THAM GIA LAP RAP VA CHUAN B| CHOLO PHAN UfNG HOAT BQNG

De khdi dpng difdc 16 phan ufng dung ngay qui dinh, can b6 VienNghien cufu Hat nhan Da Lat da lao dong hai ca trong di'eu kien thcfi tiet muabao cua cuoi nam 1983. Dae biet tham gia tich ciic trong giai doan lap rap16 va chuan bi kh6"i dong co cac can bp sau day :

/. Cac can b$ trong Ban quan ly cong trinh:

Pham Khac Chi, Trinh Hong LTnh, Le thanh Liem.Bui van Quyen, HaDang Tien,Nguyen Th| Thanh.

2. Cac can bo lanh dao va KHKT cua Vien NCHN DL:

Nguyen Mpng Sinh, Pham Quoc Trinh, Tran Lam, Nguyen vSn De,Ngo Phu Khang, Ha van Th6ng, Vo Qu6'c L^p, Nguyen Viet Long, Nguyenv&n Quy, Hoang Xuan Vu, Phan v§n Hieu, Nguyen The Ky, Le Minh Tuan,Phan Cong Thuyet, Vu Trung Hieu, Le van Sd, Phan Tri/c,Tran'T(ch Canh,Hoang v&n Nguyen, Nguyen v^n Hung, Nguyen Ba Hung, Nguyen vanLong, Vu Xuan Biet, Tr'an van Nguyen, Doan Ai Thu, .v.v.

3. Cac can b§ tCf Trung tarn ky thuat hat nhan TP. ho Chi Minh :

Nguyen Van Dat, Nguyen Mpng Hung, Vu Huy Thufc, Nguyen thi ThiT,Nguyin Quoc Blnh, Hoang v^n Hung, Le thj Phung, Nguyen Le Sdn, .v.v.

4. Cong ty xay dUng so 14 va Doan chuyen giaLien xo (cu).

Be co the khdi dpng L6 phan ufng dung tien dp, phan dong gop quantrpng thupc ve CB-CNV Cong ty xay di/ng so 14 do KS Nguyen Thiem lamGiam doc. Cong ty da tien hanh cong viec x&y l ip tCr dau n^m 1982 va hoanthanh vao giQa nam 1984, trong d6 nhufng phan viec chu yeu co lien quanden kh6i dpng 16 dadc gap rut hoan ta't. Doan chuyen gia Lien xo (cu) do KSUlitin lam tri/dng doan cung dong vai tr6 rat quan trpng trong viec xay l ip ,di/a vao khdi dpng va van hanh L6 phan ufng.

VN9700002

BAO CAOKH(3l DONG LO PHAN L/NG HAT NHAN DA LAT

1

KHdl DONG VAT LY LO PHAN LfNG HAT NHAN DA LATVCfl CAU HlNH VUNG HOAT KHONG CO BAY NCfTRON

(30/10/1983-1/11/1983)

Pham Duy Hien, Ngo Quang Huy, Vu Hai Long, Tran Khanh Mai

I- TiNH HiNH CHUAN Bj CAC H$ CONG NGHE

1- Lap 16 phan tfng : Lap be chufa nhien lieu tti 10/8 den 30/8 valap vung hoat 16 phan ufng tti 30/8 den 28/10/1983.

2- Ntfdc cat : San xua't nude cat du so IU0ng (khoang 30m3) deri/a thung 16 va he nude v6ng 1, chat \\JQng nude bao dam. Lay be chufanhien lieu lam be chufa nirdc ca't trung gian va su1 dung he th6'ng IQC nudede loc nu'dc 6* be nay.

3- H# ni/dc vong 1 va 2 : L ip xong he nude v6ng 1 va vong 2ngay 29/10. Cdng trlnh s6 7 (thap lam ngu6i) chua lap go xong, nhungkhong anh htfdng de*'n viec khdi d6ng vat ly cua 16 phan ufng. Tuy nhienphia Lien xo yeu c'au chaa cho he nude v6ng 1 chay vi so dudng 6'ng vabinh trao doi nhiet gay ban cho nude trong 16. Toan b6 he ni/dc v6ng 1va vong 2 se cho hoat dpng ngay sau khi khdi d6ng 16.

4- H# thong SUZ va AKNP : Cac phan sau day cua he SUZ vaAKNP hoat dong dUgic

- 6 thanh AZ va KC;- Cac bubng ion hoa Dl va DP;- Cac thiet bj hien hinh BVK-14 va hien so BVX-37 chl thj chu ky

va cong suat trong cac giai Dl va DP, dat mufc bao dpng va SL/ co';- Dap 16 khi su" co theo cac thong tin tC/ cac detecto Dl va DP va

theo cac thong tin tC/ cac he thong phg dong (tu" 3 6ng dem SNM-11 va2 buong ion hoa KNK-56) qua cac loi vao c6n nghe cua sd do logic AZ;

- Dieu khien b&ng tay cac thanh KC va AZ; phan chaa hoat dongcan lull y la chaa dfeu khien dugfc thanh AR.

5- H# thong di#n : Ba di/a dien cao the 6,6 kV vao bien the mdi1000 kVA ngay 21/10/83. He thong dien mdi cua Nha 1 dua vao hoatdong on dinh; dien thi cdng van dung bien the cu 150 kVA.

6- Hf thong gio : He thong gid Nha 1 hoat dong tot, cho chaythifflng xuyen cac he th6i P1, P2, P-3, P5 va cac he hut V2, V5, V6.6ng thai khi cdng trlnh sd' 20 da dung xong ngay 20/10/83.

7- H# do Ifeu phong xg va ifeu ke ca nhan : 0a dat xong 11 daudo gamma va tu he may do tiidng ufng trong phong dieu khien P.128 (8dau do trong Nha 16, 1 dau do ph6ng 128, 1 dau dd phong hut IQC khiP. 127 va 1 dau do nude vdng 2 tai hanh lang kfnh). Chuan bi s i n m£y dolieu gamma xach tay va cac may do beta va anpha tai loi vao ph6ng 16.Chuan bi du but do lieu va b6 do lieu ca nhan LTD va IFKU. Xong heRKS-2 nhang chaa cho lam viec.

8- H# thong KIP : Trong s6 cac he may do kiem tra (KIP)chuan bi xong he thong do nhiet d6 trong thung Id, nhi^t do loi ra binhtrao doi nhiet, ap suat ni/dc vdng 1 va 2, ap sua't cac he thd'ng IQC nu"dc Idva nude be chufa nhi^n lieu, mufc nadc trong thung Id.

9- H# thong thong tin va cufu h6a : ChJa xong he dien thoai noibo va bao hda. Dung he bo dam INTERCOM de lien he hai chieu giGaphdng dfeu khien va mat Id. Su" dgng cac binh cufu hda.

10- H# thu ntfdc phong xa : Sir dgng difdng d'ng K3 da xong dedUa nude nhiem xa xuong chufa trong bd'n binh chufa d tang ham Nha 2khi can thiet.

11- HE THONG PHU OQNG DUNG TRONG KH6l DQNG LO

He thd'ng phg dong phgc vg khdi ddng Id gdm Ccic detecto ndtrondat tai cac kenh ngang so 1, so 2 va so 4 va cac thiet bi ghi do tu"dng Ofng(chl thi cdng suat P va chu ky T, chl thi dem cung nhu1 cung cap cac tinhieu sg" co AZ va canh bao PS). So do ghdp nd'i nhif sau :

- Tai kenh ngang so 1

KNK-56 •->"- BUE2-15- ZPT2-01Dbng hb chl thi P & T

Cac tfn hieu AZ va PS

SNM-11 • PU2-1 : SPU1-1MHe chl thi dem

Cac tfn hieu AZ va PS

- Tai kenh ngang so 2

KNK-56 BUE2-15 ZPT2-01'Dong ho chl thi P & T

Cac tfn hieu AZ va PS

SNM-11 PU2-1 SPU-1MHe chl thj demLoaCac tfn hieu AZ va PS

UIM-2

- Tai kenh ngang so 4

Nguon nuoi 500V He chuyen mach

KNK-56 KSPV-4

SNM-11 ->— PU2-1 ••->—SPU1-1M • ->—He ch( thj dem

Cac tfn hieu PS va AZ

UIM-2

tang

Chu thich : Ky hieu dung trong c&c sd db tren nhu sau :

1. KNK-56 : Bubng ion hoaDp nhay vdi ndtron cham 1.10"12 A/(ndtron/cm2/sec),Dp nhay vdi bufc xa Gamma 1.10"11 A,Dong lam viec danh dinh 250 micro A.

2. BUE2-15 : Tien khuyech dai dong logarit.

3. ZPT2-01 : May do cong suat theo thang logarit va do toe dp

cong suat. Tren mat may co cac dbng h*6 chl thi congsuat P va chu ky ta"ng cong. suat T.

4. SNM-11 : 6ng dem ndtronDp nhay y = 1,2 xung/(notron/cm2),Phong gamma = 1 xung / phut khi D = 1000 R/h

5. PU2-1 : Tien khuyech dai cho 6'ng dem SNM-11.

6. SPU1-1M : May dem ndtronMay co : - Chi thi so dem theo he cd so 2 va dong hb dem cd hoc

(dbng ho noi tiep vdi he dem cd so 2 vdi kha n&ng chpn he so dem K ).- Loi ra cho cac tin hieu canh bao va su" co

PS = (15 den 25)K/y va AZ = (30 den 50)K/y.Cac tin hieu AZ tu" cac thiet bi phu dpng dUdc mlc noi tiep vao he

logic AZ cua he thong SUZ 6 cho gianh cho cac tfn hieu cdng nghe.

7. UIM-2 : May do toe dp dem ndtron, chl thi dbng hb xung/s.

8. KSPV-4 : May tu ghi d6ng t i bubng ion hoa.Thang do dieu chlnh dUdc nhd he chuyen mach.

Vi trf chuyen machDdng do (nano A)

10.2

20.3

30.6

41

52

65

710

820

950

10100

111000

(thang do co 100 vach va gia" tri dbng do la gia tri 6 vach thur 100).

9. Cac kenh ghi ndtron ba"ng 6'ng dem SNM-11 noi vdi may demSPU1-1M dung de do so ndtron N1 (kenh 1),N2 (kenh 2) va N3 (kenh 4).

TO cac so lieu nay ve cac dUdng cong 1/N de ngoai suy khoi lUdng tdihan.

Ngoai ra trong qua trinh nap nhien lieu can theo doi :- May ta ghi KSPV-4 ghi ddng cua buong ion hda KNK-56 cho

biet dd tang cong suat tadng doi.- Cac chf thi cong suat P va chu ky T tren cac may ZPT2-01.- Cac so dem chl thi cdng suat tren cac may UIM-2.- Chi th| cong suat va chu ky theo cac kenh Dl va DP tren man

hinh va bp chl thj so lay tu1 he thong AKNP.

Ill- QUA TRiNH KHdl DQNG VAT LY

1. Dieu ki#n an toan

Day la qua trinh nap nhie'n lieu vao vung hoat de dat trang th&itdi han, trUdng hdp kh6ng c6 b ly ndtron. Thu"c hien theo "Chifdng trlnhkh6i d6ng vat ly 16 phan ufng" [1], co thay doi mot ft thOf tu nap cdc b6nhien lieu (BNL) de thuan tien v'e thao tdc.

Qua trlnh nap nhien li#u bat dau ngay 30/10/1983 vao luc 14h30va dat tdi han luc 19h50 ngay 1/11/1983.

Ngubn ndtron khdi dpng la nguon Californi-252 co cudng dp 5.107

n/s da du"dc dat vao tarn vung hoat, 6 7-6.NgUSng canh bao PS va su* c6' AZ diidc dat nha sau: doi vdi 46

BNL dau thi PS = 25 x 256/y va AZ = 50 x 256/y doi vdi ca 3 he demSPU1-1 ghiN1,N2, N3. TO BNL thuf 50 trfi di cac gia tri ngUdng di/0c chotrong bang sau:

N1N2N3

PS25 x 4096 Ix25x16384/y25 x 4096/y

AZ50 x 4096 / y50 x 16384/y50x4096/y

2. Khoi lifting tdi hgn

Thuf tu va so lUdng nhien lidu nap vao da dapc neu trong [1]. Bexet cu the qua trinh nap nhien lieu ta nh&c lai nguyen tac nap nhienlieu de dat trang thai tdi han Ian dau.

Theo [1], nhien lieu difpc nap theo tting nhdm, trong do 3 nhom d'aumoi nhom 6 BNL. Cac nhom tiep theo dtfdc xac d|nh tC/ thuc nghiem nhasau. Gpi mk la so BNL da0c nap b Ian thuf k daa tren cac dadng congthac nghiem M = Ni (0) / Ni, i=1, 2, 3 ta nhan dadc gia tri ngoai suy tat M= 0 cua khoi ladng tdi han thieu, mk(0). Khi dd theo [1] so ladng BNLnap vao fan thuf k + 1 dapc cho bdi

m(k + 1) = [mk(O) - Tong mk] / 4 (1)

Cong thtfc nay di/a theo quy pham an toan 16 phan ufng cua Lienxo.

Hinh 1 trinh bay sd do nap nhien lieu vung hoat.Bang 1 trinh bay cac ket qua N1, N2, N3 theo thuf tu va so lUdng

BNL dUdc nap vao vung hoat va theo vi trf cac thanh dfeu khien (2 thanhAZ luon rut kh6i vung hoat, vj trf cua 4 thanh KC dUdc ky hieu :

0 = thanh rut kh6i vung hoat,1 = thanh nam hoan toan trong vung hoat).

B6'i vcfi hai may dem xung UIM-2 mac 6" loi ra cua cac may SPU1-1M tu" hai kenh ngang so 2 va so 4, c&c gi& tri sua't dem deu gan trungvdi cac gia tri ghi diidc tCf cac may dem SPU1-1M tUdng ufng. Mot so vidu du"dc neu ra, trong ngoac ( ), 6 Bang 1 cho cac fan nap nhien lieu tu"thuf 9 de thuf 12.

Tr§n Bang 1 c6n trinh bay cdc gi£ tri. cfing su^t cua 16 nhan difdctren man h)nh BVK-14 doi vcfi cac bu'dng ion h6a DM, DI2, DI3, DP1,DP2, DP3 va gia tri d6ng cua cua bubng ion h6a KNK-56 nhan dUdc trenmay \U ghi KSPV-4.

Cac hinh 2 va 3 trinh bay c£c dudng cong Ni (0) / Ni phu thuoc vaoso BNL da nap va cac gid tri ngoai suy cua khoi !i/0ng tdi han.

Trong 3 he dem ndtron phu d6ng, 6'ng dem SNM-11 dat tai kenhthuf nhat xa vung hoat nen ft nhay hdn 66i vdi si/ khong doi xu'ng khinap nhien lieu, do do dtfong cong N1(0) / N1 giam ddn dieu hdn haidudng cong 2 va 3.

Bang 2 trinh bay cac th6ng s6 trong qu^ trinh nap nhien lieu nhtfso BNL nap vao cua m6i nhom, mk, gi^ tri ngoai suy toi thieu cua khoiIU0ng tdi han sau khi nap nh6m thOf k, mk(0), t6ng so cac BNL da nap,va so lUdng BNL c6 the dUdc nap trong nhom thuf k+1, m(k + 1), di/a theocong thufc (1).

TO cac so lieu cua Bang 2 ta thay r&ng s6' laong BNL dUdc napthi/c te Idn hdn 2-3 Ian so vdi so liidng tfnh dUdc theo cong'thufc (1).Can lull y r^ng viec nap nhien lieu nhu vay la khdng th6a man yeu cauv'e an toan 16 phan ufng. Tuy nhien do than trong trong qua trinh napnhien lieu nen da khong xay ra trudng hdp dap 16 vl sii c6.

3. Qua trinh nap nhien li$u

Trong qua trinh nap nhien lieu co cdc giai phdip ky thu^t da difdcth6a thuan la :

10

Hinhl. So* do nap nhien lifu vao vung hoat.

1 .1 , ... ,2.1 ,...

A1 , A2

K1, K2, K3, K4

€'\

ern

Cac 5 1 .1 ,,.. ,2.1 ,•••

Cac bo nhien li§u mang s5 161,174,...

Cac thanh bao vf sir co AZ1 , AZ2

Cac thanh bu trur KC1 , KC2, KC3, KC4

Thanh dieu khien t\x dong AR

Thanh chen Berili

Thanh chen nhom

B*y no'tron

11

Bang 1 - Su1 phu thuoc cua cac thong so khcfi dong 16 vao khoi lu"0ngnhien lieu nap vao vung hoat.

1 - Xem Hinh 1, 2 - Vi tri KC : 0 = rut, 1 = nhung, 3 - thdi gian 1 fan do :phut, giay, 4 - suat dem trung binh cua 3 Ian do, 5 - cong suat cho tren man hinh,6 -dong buong ion hoa tren may tu" ghi.

TT

SoBNL

0

01

...

6

2__.

12

3...

18

4...

245

—306

—367...418

...

46

Vi trinap

1Nguon Cf

7 -66-6 ,8-7 ,

7 - 5, 7-76-7, 8-65-6 ,9-6 ,6-5, 8-88-5, 6-85-5,9-7,7-4, 7-89-5, 5-74-4, 4-54-6, 4-710-4,-510-6,1079-8, 8-96-9, 8-55-4, 6-47-3, 8-49-4. 7-95-3, 6-37-2, 8-39-35-9, 6-107-10, 9-98-10

VtKC

21

01

0

1

0

1

0

10

10

10

10

10

Tgdoph

35

51401

401

401

401

401

4011

11

11

11

11

SuatN1

18

1928

32

38

43

46

54

5770

6786

94125

131193

142220

dem,N2

453

5789

99

126

143

154

182

209255

233300

315433

424629

458720

xung/sN3

50

5485

95

122

135

164

186

181215

256324

294401

318450

445694

CongDM

/ Gia tri

1.3E-5

2.5E-5

3.2E-5

4.0E-5

5.0E-5

5.0E-5

6. 3E-57. 9E-5

7.9E-51.0E-4

1.0E-41.0E-4

1.0E-41.6E-4

1.6E-4<2.5E-4

suat, %N,DI2 DI3

5Ni (0)/ ~

3E-6 5E-6

6E-6 1E-5

1E-5 1E-5

8E-6 2E-5

8E-6 2E-5

1E-5 2E-5

2E-5 2E-52E-5 3E-5

2E-5 3E-53E-5 5E-5

3E-5 4E-53E-5 5E-5

3E-5 6E-55E-5 1E-4

IE-5 6E-56E-5 1E-4

KSPV4nA

6

0.051

0.053

0.054

0.055

0.056

0.057

0.0560.057

0.0580.060

0.0620.064

0.0620.064

0.0680.066

12

9. . .

50

10---54

11. . .

58

12. . .

61

13. . .6414...

66

15...

68

16. . .

69

17. . .

69

3-3,3-4,3-6,3-711-3,11-4,11-6,11-72-4,2-5,12-5,4-2

3-2,10-2,3-8

11-2, 1-810-9

4 - 9

5- 10

5-2 ,9 -2Rut

vavava

9 - 10

1

0

1

0

1

0

1

0

10

1cac

0

1cac

KC1KC2KC3KC4

1BNL dtfOc

Rutvava

Thanh6 7-1

TOl

KC1KC2KC3Bel

HAN

1

1

1

1

1

1

1

1

1

11

1BNL

1

1BNL

1111

1dtfa

1111

154

265

182

364

214

513

256

796

2941246

3466U0c

1917

3686\J0c48273414464654

385

577(570)1027

(1050649(640)1276

(1300)839

(820)2058

(2050)1018

(1000)3203

(3000)11144688

1200dtfa7259

1377

498(480)

850> (800)

594(580)1161

(1050)686

(640)1627

(1400)767

(720)2338

(1800)950

3942

1064

2.0E-4

3.2E-4

2.0E-4

4.0E-4

2.5E-4(2.5E-35.0E-4(6.3E-32.5E-4(3.2E-37.9E-4(1.0E-23.2E-41.3E-3(1.3E-24.0E-4

5E-5

8E-5

5E-5

1E-4

5E-4

1E-3

6E-4

2E-3

3E-3;

vao ttfng bade6406

1144

2.5E-3(2.5E-24.0E-4

5E-3

dtfa vao tCrng btfdc18773208547916543

1434

155321163874

13699

1246vao ttfng bade4958021655401

1928346562981502

1741264547271262

5.0E-3(5.0E-24.0E(4.0E-3

4.0E-2(40E-1

2E-2

1E-3

8E-2

KHI KC1 = KC2 = KC3 = 0 CM VA KC4 =NHUNGVAO VUNG HOAT

8E-5

2E-4

1E-4

2E-4

1E-44E-3)3E-4

1E-2)2E-4

5E-3)4E-4

2E-2_L2E-46E-4

3E-2)2E-4

1E-34E-2)2E-4

3E-31E-1)2E-4

8E-3)

2E-26E-1)

6 CM

0

0

0

0

0

0

0

0

0.0.

0.

0.

0.

0.

0.

0.

.071

.080

.073

.085

.075

095

077

110

076132

080

166

082

200

080

74

13

N\

Kenh ngang so 1

Kenh ngang s5 2

Kenh ngang so 4

I) 2 <1 16 M8 SZ ft So 6H

BN2. Six phu thuoc cua t i so N(o)/Ni vao so BNL da nap

va cac gia t r^ ngoai suy cua khoi lirong tc?i han

U

1 .0 .- 1 .00

0.'-)

0 . 8

0.7 i

D.6 _

5.5

N i °

V— / ' / A'

+o

i7id.7li

i2 •; A Kenh so 1

o Kenh 36 2

f Kenh so 4

1 L-!

10 20 30 40 50 60 BNL

Hinh 3* Sir phu thuoc cua cac h§ so nhan du-o'i to'i han

vao 36 lirqug BNL ciirq'c nap.

15

Bang 2 - Khoi Ii/Ong tdi han trong qua trinh nap nhien lieu.1- So BNL nap Ian thuf k2- Khoi lUOng tdi han ngoai suy sau Ian nap k3- Tong BNL sau Ian nap thuf k4- So BNL nap 6U0C Ian thuf k + 15- So BNL thi/c nap Ian thuf k + 1

TT

12345678910111213141516

mk16666665544433221

mk (0)2142532384249495460

64.564.566.268.369

69.5-

Tong mk36121824303641465054586164666869

mk + 142

3.23.53.53

3.222

2.52.51.51.51

0.70.4

-

mk + 1 thtfc5666665544433221-

16

- Thay d6i thuf tu* nap nhi§n lieu so vdi hudng dan [1] vi ly do thuantien trong thao tac. Nhtf v£y nhien li#u da khong d\JQc nap mdt cdch d6'ixufng vao vung hoat. Do do s6 dem cua ba he dem ndtron phu d6ngSNM-11/SPU-1M thay d6i khdng ddn dieu va cac dUdng cong Ni (0) / Nithang giang Idn (cac phep do deu dUdc l ip lai 3 fan, to/ 6* fan nap cuoi,Ian thuf 16 va 17 khi da dat tdi han).

Ket qua la gia tri ngoai suy toi thieu sau moi fan nap cua khoilUdng tdi han nh6 hdn gia tri nh^n dU0c trong tradng hop nap doi xufng.

- So lu"0ng BNL trong m6i Ian nap nhi&n lieu nhi'eu hdn so vdicong thufc (1). Dieu nay cd the chap nhan 6\J0c tti cac l$p luan sau:

+ Do viec nap nhien li#u khong doi xufng nen gia tri ngoai suy toithieu cua khoi lUOng tdi han nh6 hdn khi nap doi xufng;

+ Theo tinh toan ly thuyet, khd'i ladng tdi han la 77 BNL Do doco the coi so lU0ng BNL tdi han khong nh6 hdn 60.

That ra viec thu"a nhan qua tr)nh nap nhien lieu nhif vay la khongdung nguyen tac an toan, song moi Ian nap nhien lieu d'eu xem xet canthan viec tang so dem ndtron tu" cac 6'ng dem SNM-11 va trong cac fannap cuoi cung cd stf dung quy che nap tC/ng bade.

Mat khac, nap khong doi xufng trong moi fan nap, song phai napsao cho cac vong BNL bao quanh ngubn Cf - 252 (6 7-6) 6U0c nap daytti trong ra ngoai. Do do, sau khi nap day moi vong, phan bo nhien lieutrong vung hoat cung doi xufng.

Vong thuf nhat 6 BNL, vong thuf hai 12 BNL, v6ng thtf ba 16 BNL,vong thuf ta 18 BNL, v6ng thuf nam 29 BNL, vong thuf sau 17 BNL.

Qua trinh nap nhien lieu eg the duXfc tien hanh nhu sau :

- V6ng thuf nhat va thuf hai dUdc nap lam 3 fan, moi fan 6 BNLtheo dung hudng dan [1],

- Vong thuf ba dUdc nap lam 3 Ian, moi Ian 6 BNL (cac nhdm thuf4,5,6 d Bang 1) nhi/ng khong theo thuf tu* cua hi/dng dan. Nhdm thuf 4gbm cac BNL gan kenh ngang s6 2, do do so dem N2 tang nhanh, vlvay tren hinh 2, difdng cong N2 (0) / N2 giam nhanh hdn dtfcfng congN3(0)/N3. Nhdm thuf 5 gbm cac BNL n&m gan kenh ngang so 4, do dodudng cong N3(0) / N3 giam nhanh. Nhdm thuf 6 gbm cac BNL gan kenhngang so 1 va so 2, do do N2(0) / N2 giam nhanh hdn N3(0) / N3. TO dothi Ni(0) / Ni ta thay r&ng gia tri ngoai suy td'i thieu khoi Ii/dng tdi hantrong cac trirdng hdp nay nh6 hdn trUdng hdp nap doi xufng.

- V6ng thuf tir dadc nap 4 Ian (cac nhdm 7, 8, 9, 10). Nhdm thuf 7gbm 5 BNL nam giufa cac thanh KC2 va KC3, nhdm thuf 8 gbm 5 BNL tif

17

KC1 den KC4. Nhom thtf 9 gom 4 BNL nam hai ben thanh AZ1. NhomthuMOgbm 4 BNL na"m hai ben thanh AZ2. Doi vdi v6ng nay tfnh chatbat doi xufng trong viec nap nhien lieu cung the hien tren cac dadngcong Ni(0) / Ni va do do nhan dagfc gia trj ngoai suy toi thieu nh6 hdncua khoi ladng tcfi han. 6 cac Ian nap nay cung da giam so ludng BNLtrong m6i nhom.

- Vong thuf nam di/0c nap 6 Ian (cac nh6m 11 den 16). Vj tri napcac BNL dadc chon sao cho cac thanh AZ va KC dadc bao boc kin traderoi sau 66 mdi den cac vi trf khac. So lu"0ng BNL nap trong moi nhomdUdc gi&m dan tC/ 4 xuong 3 rbi 2 va 1. Bon thanh KC va thanh AZ1di/dc boc kin, quanh thanh AZ2 c6n thieu 6 12-4 vi 6 nay nam giufathanh AZ2 va ong chuyen mau b^ng khf nen, cay sao co di/dng kfnh 40mm de g ip nhien lieu khong dUa vao lot. Ke tu1 nhom thuf 14, cac BNLdUdc daa vao vung hoat ti/ng bade (4 den 5 bade) nho" cd cau dieu khiencan cau.

- Sau khi daa vao BNL thuf 69 va rut ca bon thanh KC th) nhanthay 16 phan ufng gan dat trang thdi tdi han : chu ky tSng cdng suat 16giam xuong dadi 100 sec. Tuy nhien, 16 chaa dat tdi han bd\ vi khi rutngubn ndtron Cf-252 ra kh6i vung hoat thi khong duy tri dadc cong suat.

- Tiep tuc daa them mot thanh Berili vao 6 7-1. Rut ba thanhKC1 - KC3 ra kh6i vung hoat. Rut dan thanh KC4 len va khi KC4 lengan net thi chu ky tang cong suat dat gid tri dadi 100 sec. Tiep theo,rut ngubn Cf-252 ra dbng thefi daa dan thanh KC4 len, khi thanh KC4con nhung vao vung hoat c9 6 cm th] dat dadc trang thai tdi han, congsuat Id phan ufng kh6ng thay doi :

19 GlCf 50 PHUT NGAY 1 THANG 11 NAM 1983.

IV- MQT SO NHAN XET KET LUAN

1.TO cac db th| Ni(0) / Ni, Hinh 2, ta thay r£ng cac da6ng congd'eu co dang lorn, nghTa la dang cho gia tri ngoai suy cua khoi ladng tdihan nh6 hdn gia tri khoi ladng tdi han thac. Di'eu nay bao dam an toan khinap nhien lieu.

2. Tren hinh 3 trinh bay su* phg thupc cua cac he so nhan daditdi han Ki = 1 - Ni(0) / Ni vao so ladng BNL dadc nap. He so nhan Ktrung binh dat gia tri 0,98 vdi 61 BNL va bang 0,99 vdi 66 BNL. Sau khinap 68 BNL, he so nay la K = 0,995, tufc la do sau dadi tcfi han dk = 1- K= 0,005 < p , vcfi p = 0,0064 la phan dong gop tadng doi cua ndtron tre.

18

3. Khi trong vung hoat chl c6 ngubn Cf-252, so dem tai kenh so 1la N1(0) = 19,4 c6n sau khi nap 68 BNL th) N1 = 4654, nhu vay congsuat 16 da tang N1/N1(0) = 240 Ian. He so tang cong suat nhan dUdc tC/cac so lieu Bang 1 doi vdi cac vi tri do N1 - N3 va DM - DI3 dUdc chodu'di day :

Vi triHe so

N1240

N2290

N3255

DI1365

DI2-

DI3500

Nhu' vay sau khi nap 68 BNL cong suat 16 da tang khoang 2 bac sovdi cong suat ngubn ban dau. Gia trj thuc cua cong suat chi/a xac dinhdi/dc bdi vi thang do cong sua't cua he AKNP chi/a dadc chuan djnh. Cacgia tri cong sua't cua DI1-DI3 ti'nh ra % cong suat danh dinh N cho trongBang 1 chl la gia tri quy Jdc de cho biet dp tang cong suat tUdng doi mathoi. Khi mdi dat ngubn Cf-252 va chua c6 nhien lieu, cong suat 16 chlvaoc5 1.10"8 %N, chuf khong phai 1.E"5 %N nhu da neu trong Bang 1.

Tuy nhien, tu" cac gi6 tri d6ng cua bubng ion hoa KNK-56 dat taikenh ngang so 4 va chl thj tren may \\J ghi KSPV-4 thi he so tang tUdngdoi cua cong suat la 0,2/0,05 = 4 fan : gia tri nay be hdn nhieu gia trj daneu tren day.

4. Theo tfnh toan ly thuyet, 16 phan ufng dat tcfi han vdi cau hinhkhong ,co bay ndtron khi vung hoat di/dc nap 77 BNL (xem hinh 3 cua[2]). Nhu'ng ket qua thUc nghiem da cho thay 16 dat tdi han khi nap vaovung hoat 69 BNL va 1 thanh Berili. NhU vay ket qua tfnh toan ly thuyetkha I6n hdn gia tri thiic nghiem.

Tuy nhien trong tinh toan ly thuyet, khoi lUdng U-235 trong moiBNL dude cho bang 38,9 g, trong khi 66 khoi lUdng U-235 thuc te trongmoi BNL la 2781,8/69 = 40,3g. Bang 3 trinh bay cac thong so ve cacBNL (so hieu nha may va khoi lUdng) theo thu" tu nap nhien lieu. Ta thayrang khoi liidng tong cpng U-235 nap la 2781,8 g, tUdng Jng vdi soludng BNL "ly thuyet11 bang 2781,8/38,9 = 71,5 BNL Tuy vay, so UgngBNL ly thuyet co the Idn hdn 71 thanh b6i vi theo nhu Bang 3, khoi UdngU-235 trung binh cua cac BNL 6 vong nhien lieu trong cung, la vongBNL co trpng so Idn nhat, la 40,9g, c6n cac vong nhien lieu tiep theo cokhoi lUdng trung binh la 40,2 g, 40,44 g va 40,3 g.

19

BangJ3- Cau hinh tdi nan 69 BNL, thtf tu" nap nhien lieu va khoi lu"0ng U-235

Thu1 TVVong1

2

3

Nhom1

2

3

4

5

6

NapSo123456789101112131415161718192021222324252627282930313233343536

Vi tri6-68-77-57-76-78-65-69-66-58-88-56.85-59-77-47-89-55-74-44-54-64-710-410-510-610-79-88-96-95-85-46-47-38-49-47-9

BoSo hieu056057058059060061068069070074075076077078079086087088062063064065066067080081082083084085071072073155156157

nhien lieuKhoi Ii/Ong U-235, g41.140.840.541.241.140.739.738.240.740.841.741.141.339.240.538.640.440.840.238.639.839.841.040.039.239.340.341.541.040.240.739.240.039.840.339.9

T = 245.4m = 40.9

T = 483m = 40.3

T = 720.8m = 40.04

20

4

5

7

8

9

10

11

12

13

14

15

16

373839404142434445464748495051525354

555657585960616263646566676869

5-36-37-28-39-35-9

6-107-108-109-93-33-43-63-711-311-411-611-72-42-512-54-23-210-23-811-210-911-84-95-105-29-2

9-107-1

158159160164165166170171172176177178179180181182183184161162163173174185187175167168169168189190191

Berili

40.040.741.840.541.141.441.340.739.639.940.039.140.240.241.838.340.140.840.740.841.340.241.039.641.139.940.540.138.840.540.140.239.7

T = 727.5m s 40.44

T = 604.5m = 40.3

Khdi Ii/Ong U-235 tong c0ng tdi h$n T = 2781.8 gKhoi ItfOng U-235 trung binh mot BNL m = 40.3 g

21

V. THANH PHAN NHOM KHdl DQNG LO

Nhom khdi dpng 16 g'om cac thanh vien va nhiem vu cua ho nhu" sau.

TT123

45

6

78

9

10

11

12

Chu'c vuPhu trach kh6i dpng

Trud'ng kipVat ly kiem tra

Ks dieu khienKs thiet bj phu dpng

Ks he SUZ

Ks he kipKs cd khi

Ks dien

Ks an toan biJc xa

Ks vat ly

Ks che dp nadc

Phfa Lien xoArhanghenski N.VArhanghenski N.V

Terekhov A.V

Kuznexov E.MTrophimov A.BLavrenov N.PKharlamov N.B

Sokolov I.VBakharev V.S

Truskin V.S

Khandamirov U.E

Semennov S.M

Phia Viet namPham Duy HienNgo Quang Huy

Vu Hai LongNguyen Oang Nhuan

Tran Khanh MaiNguyen Ngpc Lam

Nguyen Nhj Dfen

Tran Khac AnDo van Hiep

Pham Xuan PhUdngNguyen Trudng Sinh

Pham van LamTrUdng Cam Ranh

DUdng Quang TanNguyen Tac Anh

LUdng Ngpc ChauHuynh Dong PhUdng

Le VTnh VinhNguyen thi NangTC/ van NghTa

VI. TAI LIEU THAM KHAO

1. Chiidng trinh kh6i dpng vat ly 16 phan ufng,Tai lieu Thiet ke Lien xo, 1980.

2. Qui pham bao dam an toan khfti dpng vat ly 16 phan ifng,Tai lieu Van hanh va khai thac 16, Lien xo, 1980.

22

IllVN9700003

B A O C A OD0NG LO PHAN L/NG HAT NHAN DA LAT

KH6l DONG VAT LYVC3l CAU HJNH VUNG HOAT CO BAY N0TRON

(15/12/1983-27/1/1984)

Pham Duy Hien, Ngo Quang Huy, Vu Hai Long, Tran Khanh Mai

I - TINH HlNH CHUAN B| BE KHdl DQNG

L6 phan ufng dadc nap nhien li£u Ian dau theo cau hinh khong co bay ndtronva dat tdi han ngay 1/11/1983 vdi 69 BISIL (xem/1/, cung vdfi danh sach tham gia khdidpng).

1. So vcfi fan khdi dpng trtfo'c. Cac h# thong cong nghe, da dUQc chuan b|them trud'c khi nap nhien lieu g*6m co :

- He nude tuan hoan vong 1 va vong 2 : Da hoat dpng 6" che dp danh dinh.Thap lam ngupi da lap xong gian go nhUng cht/a che chan xung quanh.

- He di'eu khien SUZ va AKNP : day du, ke ca thanh AR.- He may do kiem tra KIP: Cac may do li/u li/dng ni/dc vong 1 va vong 2 da

hoat dpng.- He thong dien : He thong dien su" co sCr dung diezen da hoat dpng.- He thong thoat ni/dc nhiem xa : Noi he K3 tarn den Nha 5 trong do SL/ dyng

dudng K3 tCr Nha 1 den Nha 2 va di/dng di/a ni/dc ngi/dc tu1 Nha 5 den Nha 2.Cac he do dem phg dpng de khdi dpng cung giong nhii fan tri/dc (xem tai lieu

ni).

2. Sau khdi dpng vat ly fan trade, ngay 5/11/1983 phat hien si/ xam mot thanhnhien lieu va thanh chen bang nhdm. Hi#n ti/dng nay xay ra khong binh thi/dng choden ngay 15/12/1983. Trong qua trinh nap nhien lieu, tiep tuc theo doi hien tUdngnay de cd ket luan trade khi khdi dpng nang ladng ( xem phu luc ).

3. Ngay 15/12/1983 da xac dinh thdi gian rdi cua cac thanh dieu khien trongvung hoat :

AZ1 va AZ2 : 0,30 sec ; KC1 ; 0,30 sec, KC2 : 0,29 sec ;KC3 : 0,28 sec, KC4 : 0,26 sec,

23

II- TRANG THAI TCfl HAN VCll CAU HJNH VUNG HOAT CO BAY NCfTRON

Dtfa vao ket qua tdi han vdi 69 BNL cua fan khdi dong trade, viec nap nhienlieu fan nay co the tien hanh timg nhdm nhieu bo NBL so vdi cong thufc (xem IM).

Ngay 15/12/1983 bat dau nap nhien li#u, song gap khd khan trong thao tac dotru*dc do trong vung hoat khong dat du so thanh nhom chen.

Ngay 16/12/83 dat tat ca cac thanh nhom chen can thiet vao vung hoat va rutong chuyen mau khi nen d 6 13-2 ra. Ly do la ong nay co dUdng kfnh phan tren vunghoat bang 40 mm trong khi kich thirdc BNL la 35 mm, cho nen khi nap nhien lieu 6ke can, thanh nhien lieu co the bi va cham gay stfdt vo boc; hdn ntfa BNL 6" 6 12-4khong nap vao dtfdc vi khong gian qua chat.

Ngay 17/12/1983 lai bat dau qua trlnh nap nhien lieu va dat tdi han ngay18/12/1983 luc 17 h 48.

Thuf to" nap nhien lieu, Bang 1, gom bay fan nap. Nhom thuf nhat 48 BNL,nhom thuf hai 8 BNL, nhom thuf ba 6 BNL, nhom thuT tu 3 BNL, nhom thuf nam 3BNL, nhom thtf sau 3 BNL va nhom thi bay 1 BNL : Bgt trang thai tdi han vdi 72BNL trong cau hlnh vung hoat co bay notron d trung tarn. Trong cau.hinh nay dadung 12 thanh Berili. Hinh 1 trinh bay cau hinh nap nhien lieu va Hinh 2 trinh bay su1

phu thuQC so dem ngiidc vao khdi lu"0ng nhien Ii#u U-235 cua cac BNL dtfde chotrong Bang 1. Tong khdi lifting U-235 tdi han trong cau hinh nay la 2897,4 g.

III. NAP NHIEN LIEU CHO CAU HJNH LAM VIEC

Qua trinh nap nhien lieu de tao mQt cau hinh lam viec bj cham lai do phat hienmot so BNL bi xam be mat (xem Phu luc 1). Du"di day, vj trf cac thanh dieu khiendu"dc tinh bang mm, do phan Ofng va do hi#u dung bang ddn vj p eft ($).

Ngay 19/12/1983 do dp hi#u dung cua thanh AR dat den 1 $. Gia tri nay Idnhdn so vdi quy djnh an toan hat nhan, tu" 0,4 den 0,8 $. Do do ngay 21/12/83 da thaydoi vi trf cac BNL; di/a vao kho BNL so 067 d 6 7-10, chuyen vi trf BNL 074 ttf 6 6-11sang 6 2-7 va BNL 072 ta 6 8-1 sang 6 12-7. Ngoai ra dat them 3 BNL so 078, 180va 187 vao cac 6 12-2, 12-3 va 12-4; ba BNL nay da bi tray sUdt trong qua trinh napnhien lieu ngay 17 va 18/12/83. Cau hinh nhien Ii#u ngay 21/12/83, 74 BNL va 13thanh Berili, dUdc the hien tren Hlnh 3. Khdi li/dng tong cpng U-235 la 2977,9 g.

Cau hinh nay, vdi vi trf cac thanh dieu khien la : KC1 = KC3 = KC4 = 0 mm,KC2 = 442 mm, AR = 0, dU0c dung den ngay 26/12/83 de xac dinh dp hi#u dung cuacac thanh dieu khien va mot so thanh nhien lieu va Berili, xem Bang 18.

24

BANG 1 - Cau hinh tdi han 72 BNL, thuf tu". nap nhien lieu va khoi \U0ng U-235.

Thuf tu" nap

nhom

1so

1

2

34

567

89

1011

12131415

1617

18

19

2021

22

2324

252627

28

2930

3132

33

34

3536

Bo

Vi tri

6 - 5

5 - 5

5 - 7

5 - 6

6 - 87 - 88 - 89 - 69 - 7

9 - 57 - 4

8 - 58 - 47 - 36 - 4

5 - 44 - 4

4 - 5

4 - 6

4 - 7

5 - 8

6 - 97 - 9

8 - 9

9 - 81 0 - 71 0 - 510 -610 - 411 - 611 - 7

9 - 98 - 10

7 - 10

5 - 9

6 - 10

nhien

so hieu

177

162

161

070076158178164

075069188171173194

179

176189

191160

062064

088181063

080081174

169

065

195086

190163

067

175

082

lieu

khoi IU0n40.040.840.740.741.140.039.140.541.738.240.540.740.241.040.239.940.139.741.840.239.840.841.838.639.239.341.038.839.839.038.640.241.340.039.940.3

U-235, g

T = 1927,3 g

25

BANG 1 - (tiep)

Thuf tu*

Nhom

2

3

4

5

6

7

nap

So

37

38

39

4041

42

4344

454647

48

49

505152

5354

55

56

57

5859

60

6162

6364

65

66

67

68

69

7071

72KhoiKhoi

Vi tri3 - 73 - 63 - 43 - 35 - 36 - 37 - 28 - 39 - 49 - 311 - 411 - 31 2 - 512-611 -810- 99- 108- 115- 106- 11

4 - 9

3 - 8

2 - 6

2 - 52 - 4

2 - 3

3 - 2

4 - 25 - 2

6 - 2

8 - 2

9 - 2

1 0 - 2

11 - 27 - 1

2-2lircfng U-235 tong cItfOng U-235 trung

Bo

So hieu165167156192

193

186

066185087

182

183184

073084

071

159

172

072

083074

168157155

166

085079

068077

170059061125126127060

149:ong tdi nanbinh 1 BNL

nhien lieu

Khoi lU0ng U-235, g41.140.540.339.439.840.541.039.640.438.340.140.840.041.0.40.740.739.639.241.540.8

40.139.939.841.440.240.539.741.341.341.240.739.240.539.841.1

39.9T=2897,2 gm = 40,24 g

' T= 323,5g

T = 241.9g

T= 122.3g

T= 121.1g

T= 121.4g

tdi nan

26

!l!lLhJ- hinh vun^ hoat to'i h^n co bay no'tron

gom 72 BNL.

Cac bo nhien lifu mang so 149,68,...

Cac thanh bao vf su* c6 AZ1 , A7,2

Cac thanh bu trur KC1,...,KC4

tu Qng AR

Thanh chen Berili

Thanh chen nhom

Kenh chieu mau khi nen 13.2

no'tron

Thanh dieu khien tu*

.1,...,?. 1 ,... Cac 6 1.1,..., 2.1,.

27

'' *

;\fK

A Kenh ngang a6 1

o Kenh ngang 36 2

+ Kenh ngang a5 4

-k

\\

\

\

'»\\\ V

\

\

\

\

\ .

A

Hinh 2 • Su* phu thuoc so dem nguq'c vao khoi lirq'ng

nhien li^u trong qua trinh ctat to'i nan

cua vung ho;jt co bay no'tron ( 72

28

Hinh 3« Cau hlnh vung hoat gom 74 BNL

( 21 - 26/12/1983 )

A1 , A2

K1,.. . ,K4

Cac BNL mang s$ 149,68,...

Cac thanh bao vf sir co AZ1, AZ2

Cac thanh bu trir KC1,...,KC4

Thanh dieu khien tu1 dgng AR

Thanh chen Berili

Thanh chen nhom

Bay no'tron

29

$'r4yn/**3?&$2$$J<\v

Hinh 4. '"lau hinh vung hog.t gom 86 BNL.

( 29/12/1983 )

fJj ^Q2\ Cac bo nhien li|u mang so 149,162,

A1, A2 Cac thanh bao v§ su* co AZ1 , AZ2

K1,...,K/I Cac thanh bu tru* KC1?...t KC4

. XThanh ctieu khien ttj dqng AR

Thanh chen Berili

Thanh chen nhom

Bay no'tron

30

/ •

X

Hinh 3

($jl t\ -1 .1 '

Cau h m h vung hoat gom 83 BNL,

( 30/12/1983 )

Cac BNL mang s6 149,162,...

Cac thanh bao v§ sir co AZ1 , AZ2

Cac thanh bu trir KC1,...,KC4

Thanh dieu khien tif dgng AR

Thanh chen Berili

Thanh chen nhom

Bay nctron

31

'.'I

Hinh 6. Oau hmh vung ho at gom 88 BNL

(14 - 16/1/1984 )

Cac BNL mang so 149, 162, ...

A1? A2 Cac thanh bao vif sy co AZ1 , AZ2

K1 , ...,K4 Cac thanh bu tru- KC1, ..., KC4

Thanh dieu khien tir d§ng AR

Thanh chen Berili

Kenh thi nghifm 7-1

B*y no'tron

32

/

•X'!-' -i v

\/r^ i

Uinh 7- Cnu hinh vung ho§it gom 88 BNL

> ^ay 17/1/1984 )

Cac BNL mang so 149, 162,...

Cac thanh bao v£ sir co AZ1 , AZ2

Cac thanh bu trij KC1 , ..., KC4

Thanh dieu khien tij' dong AR

Thanh chen Berili

Bay no'tron

Kenh thi nghifm 7A

33

Dp hieu dung cua thanh AR, dU0c bao bQC bdi 3 thanh Berili va 2 thanh nhomchen, da giam xudng con 0,86 $. Thanh AR sd dTco d<? hi#u dung Idn nhu vay la doda dung cacbua bor thay vi thep khong gl nhu thiet ke neu.

Do hieu dung cac thanh KC, AZ nam trong khoang 1,96 den 2,48 $. Theothiet ke cac thanh nay co do hieu dung cd 4 $.

Ngay 27/12/83 chuyen 12 BNL vong trong cung ra ngoai cung. Cau hinh naykhong dat tdi han.

Ngay 29/12/93 nap them 12 BNL vao vong trong cung, nhu" vay ta dUOc cauhinh 86 BNL vdi 13 thanh Berili, Hinh 4 , va khdi Ii/Ong U-235 tong c<?ng la 3461,4 g.Trong cau hinh nay xung quanh thanh AR van la 3 thanh Berili va 2 thanh nhom.

Vdi cau hinh nay cung da xac djnh dd hi#u dung cua cac thanh di'eu khien vamot so BNL va thanh Berili. Ket qua, Bang 18, cho thay dd hi#u dung cua cac thanhKC Idn hdn so vdi cau hinh 74 BNL. Dae bi#t la dd hieu dung cua thanh AR dat den1,04 $ qua Idn so vdi quy dinh an toan hat nhan.

Ngay 30/12/83 thay hai BNL so 076 6 5-11 va so 158 6 9-11 bang cac thanhBerili va 6 7-1 de trong. Nhu vay ta dU0c cau hinh 83 BNL vdi 15 thanh Berili, Hinh 5.Viec tao cau hinh nay nham xac djnh do hi#u dung thanh AR khi 6\J0c bao quanhbang 5 thanh Berili va 2 thanh nhom. Ket qua cho thay do hieu dung thanh AR giamxudng con 0,85 $. Gia tr| nay co the chap nhan du"0c, tuy nhien cau hinh tro" nenkhong ddi xufng khong ti#n lo"i cho viec khai thac 16.

De xem xet kha nang thay thanh AR, ngay 31/12/83 da xac djnh do hieu dungcua mot thanh thep khong gf dudng kinh 28 mm dat tai 6 7-1 ddi xufng vdi 6 vi trithanh AR. Ket qua cho do hieu dung bang 0,58 $. Nhu" vay nen dung thep khong gllam thanh AR. Do do trudc khi tim kiem mQt cau hinh lam viec da quyet dinh thaythanh cacbua bor bang thanh thep khong gl. Do la 2 thanh dai 50 cm, dtfdng kfnh 28mm, ndi tiep nhau, cac thanh nay nguyen dUOc dung de tao trong \U0ng dat dudi cacthanh dieu khien. Cudi cung da dat dU0c cau hinh lam vi#c gdm 88 BNL, 18 thanhBerili va sti dung tu" ngay 14/1/84 den 16/1/84, Hinh 6 va Bang 18.

Ngay 17/1/84 van giuf cau hinh 88 BNL - 18 thanh Berili nhUng trong do thaycac BNL sd 059, 060 va 061, la nhufng BNL da bi xam khi de d kho chufa tarn thdi tCrngay 2/11/1983, bang cac BNL mdi sd 144 6 6-2, 130 6 5-11 va 128 6 8-2 va cungthay cac BNL bj si/dt sd 078, 180 va 187 b&ng cac BNL so 124 6 12-2, 123 6 12-3 va122 d 12-4, Hinh 7.

Tom lai, da xac djnh du*0c 3 cau hinh co bay ndtron : cau hinh co do sau duditdi han nho gdm 74 BNL - 13 thanh Berili, cau hinh gdm 86 BNL - 13 thanh Berili vacau hinh lam viec 88 BNL - 18 thanh Berili. Cau hinh lam viec nay la cau hinh cudicung trong doi khao sat khdi dong vat ly va cung dung cho khdi dong nang lu"0ng vakhai thac Id tiep theo, Bang 2.

IV - DAC TRUNG TICH PHAN VA DQ HIEU DUNG THANH AR

4.1 - Phuana phap

34

trung tfch phan cua mpt thanh dieu khien la su" phu thupc cua dp phantifng hi#u dung vao dp nhung sau cua thanh trong vung hoat. 00 hi#u dung cuathanh la dp phan ufng gay ra khi thanh dirpc nhung hoan toan vao vung hoat. DOhi#u dung di/0c xac djnh khi xac dinh di/dng dac trtfng tfch phan cua thanh. Trongdp*t khdi dpng vat ly nay xac djnh dac tri/ng tfch phan cua thanh AR bang phtfdngphap chu ky nhan doi va d&c trUng tfch phan cua cac thanh khac bang phifdng phapbu tru" vdi thanh AR.

BANG 2-Cau hinh lam viec 88 BNL.18 thanh berili, vi trf nhien li#u va khoi\U0ng U-235.

SoTT13579111315171921232527293133353739414345474951535557596163

Bo nhi§n li$uVitrf2-22-42-63- 13-33-63-84- 14-44-64-95- 15-35-55-75-95- 116-36-56-96- 117-37-87- 108-38-58-98- 11.9-29-49-69-8

Ma so149085155162192167157177189160168188193151146175130186153088076194152158185145063067125087138080

m U-23539.940.239.840.839.440.539.940.040.141.840.140.539.840.840.539.940.440.538.740.841.141.041.140.039.640.138.640.039.240.439.739.2

SoTT2468

. 10121416182022242628303234363840424446485052545658606264

Vitrf2-32-52-73-23-43-73-94-24-54-74- 105-25-45-65-85-106-26-46-86- 107-27-47-98-28-48-88- 109- 19-39-59-79-9

36 nhienMa so079166074068156165070077191062161170176154064083144179147082066150181128173148163171182139137190

lieu

m U-23540.541.440.839.740.341.140.741.339.740.240.741.339.941.239.841.539.840.240.940.341.041.641.840.340.239.541.340.738.339.540.440.2

35

BANG 2- (tiep)

SoTT656769717375777981838587

Bd nhien lieuVi trf9- 1010- 110-410- 610-911 - 111 -311 -611 -812-212-412-6

Ma so'172069065169159164184195071124122084

m U-235, g39.638.239.838.840.740.540.839.040.738.740.741.0

Khdi ladng U-235 tong <

SoTT666870727476788082848688

:png da

Bd nhienVi trf9- 1110-210-510-7

10-1011 -211 -411 -711 -912-312-512-7

Ma so'143126174081178127183086075123073072

nap la 3537,3 g

lieum U-235, g

38.840.541.039.339.139.840.138.641.740.240.039.2

Qua trinh xac djnh dac trang thanh AR nhi/ sau. Baa 16 den trang thai tdi nanvdi thanh AR hoan toan nhung trong vung hoat, vi trf 650 mm. Keo thanh AR len motdoan de dtfa vao mot do phan ting dadng r: cong sua't 16 tang. Xac dinh chu ky nhandoi cong sua't T2, di/a theo phadng trlnh gid ngatfc r = f (T2) xac dinh diTtfc phan tingdadng r tu"dng ting. Bade tiep theo,xac lap lai trang thai tdi nan vdi cong suat bandau bang each thay doi vi trf cac thanh KC dong thdi van gitf nguyen vi trf thanh AR.Sau do l&p lai qua trinh mo ta tren day, dtfa 6Q phan ting dating vao bang each keothanh AR len mot doan va tiep tuc qua trlnh cho den khi thanh AR dadc rut toan bokhdi vung hoat. Sau khi xac dinh 6Q phan ting vdi tting doan thanh AR, c<?ng cacphan ting nay Ian \\J0\ theo chieu sau de xay du"ng dadng dac trang tfch phan cuathanh.

Vi trf thanh AR dUOc xac dinh tren bang chl thi vj trf d ban dieu khien. Thdigian nhan doi cong suat dtftfc xac djnh nhd dong ho bam giay . H# thong thiet bj phudpng gbm buong ion hoa KNK - 56 dat trong kenh ngang so 4 va may ti/ ghi KSPV -4 ghi dong ty l# vdi cong suat 16.

Khi xac dinh dac trang tfch phan cua thanh AR can tranh anh hadng cuating nhiet dQ, hi#u ting cong suat va hi#u ting nhiem dOc xenon. Trong khdi dQng vatly, cac hi#u ting nay chaa xuat hien vl tien hanh 6" cong suat rat be. Trong cac phepdo, trang thai tdi han dat daoc d cong suat khoang 1 . E-5% N (ttic cd 1.E-2 W), vdiN la cong suat danh dinh bang 500 KW . &Q chu ky nhan doi d mtic cong suat 1E-4% N. Trong cac dieu ki#n nay, phadng trlnh gid nga0c su" dgng cac thong so ndtrontre tti phan hach U-235.

36

Ngoai ra, trong qua trinh xac djnh d|c trUng tfch phan thanh AR con phai chuy den anh hudng cua cac thanh hap thu gan, la cac thanh KC1 va KC4. V) vaythi/dng hay khao sat cac cau hlnh KC1 = KC4= 0 mm va cau hlnh 4 thanh dieu khienKC nhung deu nhau. Sau day la k£t qua d|ic trang tfch phan va dp hi#u dung cuathanh AR bang cacbua Bor va bang thep khong gl.

Chu y - Trong thu"c te, khi khdi dpng cung nhu" van hanh 16 sau nay, v| trf kyhieu 4 thanh KC khac vdi cac ban ve thiet ke.

4.2 - Thanh AR cacbua bor

Dae trifng tfch phan diTOc xac djnh doi vdi 4 cau hlnh nhien li#u khac nhautrong vung hoat.

a) Cau hinh75BNLva 12 thanh Berili, ngay 21/12/ 83.Cau hlnh nay giong cau hlnh trdn Hlnh 3 cpng them tai 6 7-10 dat 1 BNL thay

cho thanh Berili : nhtf vay xung quanh thanh AR dat 1 BNL, 2 thanh Be va 2 thanhnhom. Cau hinh cac thanh dieu khien la KC1 = KC3 = KC4 = 0 mm, con KC2 dunglam bu tru" de dat trang thai tdi nan. KS't qua cho trong Bang 3 trong 66 chu y rang rla do hi#u dung ttfdng Ofng vdi d$ djch sau cua thanh vao vung hoat, con trong quatrlnh do thl thanh AR dU0c rut dan len theo chieu ngu"0c lai.

BANG 3 - Dae trUng thanh AR cacbua bordr : dp phan ufng dUdng Ofng vdi chu ky nhan doi T2,

r : dp hieu dung ti/dng ufng vdi vi trf nhung cua ARre : dp hieu dung chuan ve ddn vj cua thanh AR

- a)Vj trf thanhKC2, mm

365. 385

425475548650

b)Vi trf thanh

KC2 , mm317340372405455

Cau hinh 75Vi trf thanhAR, mm

6504503502501500

Cau hlnh 74Vi trf thanhAR , mm

650450325225

0

BNL 12 thanh Be:Chu ky

T2, sec-

31.514.913.416.436.9

BNL 13 thanh Be:Chu ky

T2, sec-

37.219.320.817.4

KC1 =Dp

dr-

0.1720.2690.2840.2550.153

KC1 =Dpdr-

0.1530.2340.2230.247

KC3 =hi#u

KC3 =hi#u

KC4=0dung

r1.1330.9610.6920.4080/153

0

KC4=0dgng

r0.8570.7040.4700.247

0

re1.000.850.610.360.136

0

re1.000.830.550.29

0

37

Bang 3 - (tiep)

c) Cau hlnh 86 BNLVi trf bon

thanh KC, mm363367375385392398400

Vj trf thanhAR, mm

6505004003002001000

d) Cau hinh 83 BNLVj trf bon

thanhKC, mm355365373380383

Vj trf thanhARr mm

650450325200

0

.13 thanh Be: KC1 = KC2 =Chu ky

T2, sec-

41.417.015.028.636.5100.

• KC3 = KC4DO hi$u dungdr-

0.1420.2520.2700.1840.1550.041

15 thanh Be: KC1 = KC2 =Chu ky

T2, sec-

22.615.216.7200

r1.0440.9020.6500.3800.1960.041

0

= KC3 = KC4DQ hi#u dungdr-

0.210.270.2560.129

r0.8650.6550.2560.129

0

re1.000.860.620.360.190.04

0

re1.000.760.45

' 0.150

Trong cot cudi cung cua Bang 3, d<? hi#u dgng tong cua thanh AR (= 1.13 $)dU0c chuan ve ddn vj; dieu nay giup ta so sanh dang cac dudng dac trifng tfch phantrong cac cau hinh khac nhau, Hinh 8.

b) Cau hlnh 74 BNL va 13 thanh Berili, ngay 22/12/83.Xung quanh thanh AR dat 3 thanh Berili va 2 thanh nhom, Hinh 3. Ket qua

cho trong Bang 3b vdi 6Q hi#u dung tong cua thanh AR = 0,85.

c) Cau hinh 86 BNL va 13 thanh Berili, ngay 29/12/83.Xung quanh thanh AR van dat 3 thanh Berili va 2 thanh nhom, Hinh 4. Bon

thanh KC co cung vj trf nhung sau. Ket qua cho trong Bang 3c vdi d<? hi#u dung tongcua thanh AR = 1.04.

d) Cau hinh 83 BNL va 15 thanh Berili, ngay 30/12/83.Xung quanh thanh AR co 3 thanh Berili va 2 thanh nhom.

= 0,86, xem Bang 3d.hi#u dung tong

Hinh 8 trinh bay cac dac trifng tfch phan sau khi chuan ve dcfn vj. Y nghTa cuacac di/cfng cong nay se dUOc phan tfch so sanh vd\ cAc dac tri/ng cua thanh AR bangthep khong g[ cf phan difcfi day.

38

4-3 - Thanh AR thep khdng gJ

Nhir tren da ndi, tCr ngay 14/1/84 vung hoat co cau hlnh lam vi#p 88 BNL va18 thanh Berili vdi thanh AR bang thep khong gl. Trong cau hlnh nay da khao sat d$ctrung tfch phan cua thanh AR doi vdi cac bo trf khac nhau cua cac thanh dieu khien.

a) Ngay 14/1/84 vdi 4 thanh KC co cung vj trf cho dp hi#u dung thanh AR *0.489, xem Hinh 6 va Bang 4a.

b) Ngay 23/1/84 vdi cap thanh KC2, KC3 gitf co d|nh d v| trf nhung hoan toan,dp hi^u dung thanh AR = 0.593, xem Hlnh 7 va Bang 4b.

c) Vdi vj trf ngi/Oc lai cua cac thanh dieu khien, giuf co dinh cap thanh KC1 =KC4 = 650 mm, dp hieu dung cua thanh AR giam xuong = 0.449, Bang 4c.

Ta nhan xet rang trong 3 trirdng hdp tr6n, dp hi#u dung r cua 3 thanh ARthep khong gl phg thupc nhieu vao vj trf cap thanh KC1, KC4 : r tang khi rut dan c$pthanh nay ra khdi vung hoat- Hlnh 8 trlnh bay cac du"dng d&c tri/ng tfch phan chuannay, tuy rang moi dudng cong chl du"p*c xac dinh bang 3 diem do. Sau khi khdi dpngnang lUtfng da tien hanh xac dinh lai cac d|c tri/ng tfch phan nay cua thanh AR.

d) Ngay 28/2/84, 4 thanh KC = 450 mm, r = 0.442, Bang 4d.e) Ngay 29/2/84, KC1 = KC4 = 650 mm, r = 0.461, Bang 4ef) Ngay 1/3/84, KC2 = KC3 = 650, KC1 = KC4 = 370, r = 0.529, Bang 4f

BANG 4 - Dae trtfng thanh AR thep kh6ng gl.Cau hinh 88 BNL va 18 thanh Berili.

KC1 - KC4450457467472

KC1 = KC4368375392

-

KC2 = KC3375380395

-

a) Vj trf cac thanh KC1 = KC2 = KC3 = KC4AR650400200

0

T2, s-

32.61962

dr-

0.1670.2180.104

b ) Vj trf cac thanh KC2 = KC3AR

650400250

0

T2, s-

29.517.931.9

d r-

0.1790.2450.169

c) Vj trf cac thanh KC1 = KC4 =AR

650400200

0

T2,s-

57.318.379

d r-

0.110.0.2440.085

d. h. d r0.4890.3220.104

0= 650

d . h. d r0.5930.4140.169

0= 650

d. h. d r0.4390.3290.085

0

r chuan1.000.660.21

0

r chuan1.000.700.29

0

r chuan1.000.750.19

0

39

Bang 4 - (tiep)

KC1 - KC4450

----

450----

KC2 va KC3363370385

-363372377

380, 390382, 395

KC1 va KC4360368

381, 375387

-358367

-385, 383

-

d) Vj trf cac thanh KC1 = KC2 =

AR650400300200

0650400300200

0

T2, s-

39.253.665.390.5

-37

50.356.672

e) Vi trf cac thanh

AR650400200

0650400300200

0

T2, s-

561766-

64.4. 65.5

6684

f) Vj trf cac thanh

AR650400300200

0650400300200

0

T2, s-

32.535.552

80.6-

35.834.249.756.2

d r-

0.1470.1160.1000.079

-0.1480.1220.1120.092

KC3 = KC4

d. h. d r0.4420.2950.1790.079

00.4740.3260.2040.092

0

KC1 = KC4 = 650

d r-

0.1120.2510.098

-0.1000.0990.0980.082

KC2 = KC3

dr-

0.1680.1580.1190.084

-0.1580.1620.1230.112

d. h. d r0.4610.3490.098

00.3790.2790.1800.082

0= 650

d. h. d r0.5290.3610.2030.084

00.5550.3970.2350.112

0

r chuan1.000.670.410.18

01.000.690.430.19

0

'r chuan1.000.760.212

01.000.740.490,22

0

r chuan1.000.690.390.16

01.000.710.420.201

0

40

Theo bang nay thi trang thai tdi han vdi cau hinh 88 BNL dat dUdc vdi vj trfcac thanh dieu khien nhu* sau : 4KC » 462 mm; AR = 300 mm.

4.4 - Nhan xet ve cac dac trUna tfch phan cua thanh AR

Hinh 8 trinh bay tat ca cac ket qua do du"dc cua cac dac trang tfch phan cuathanh AR bang cacbua bor va bang thep khong gl doi vdi cac vi trf khac nhau cua caccap thanh KC1 va KC4 : 650 mm, 450 mm va 360-380 mm. TCf hinh nay ta c6 cacnhan xet sau :

a) Tat ca cac dudng d$c tri/ng tfch phan c6 dang giong nhau. Chung deu cdphan tuyen tfnh trong khoang tu" 200 den 400 mm va dp hieu dung trong khoang naychiem 50% dp hi#u dung toan thanh AR.

b) Vj trf cua cap thanh KC1 va KC4 nam gan thanh AR co anh hudng manhden dang dac tri/ng tfch phan va dp hi#u dung tong cua thanh AR.

- Khi ca hai thanh nay nam ngoai vung hoat, KC1=KC4=0, (cac diem 1 va 2)hoac khi ca hai deu nam hoan toan trong vung hoat, KC1=KC4=650, (cac diem7,10,11), dac trt/ng tfch phan co dang d6'i xtfng qua vj trf 300 mm.

- Khi cap thanh KC1-KC4 6" cac vi trf trung gian, 450 mm (cac diem 5,8,9) va360-380 mm (cac diem 6,12,13) doi vdi thanh thep khong gl, thl dac trtfng tfch phanco dang khong doi xufng. Cac ket qua do khong cho phep phan biet cac trt/dng hdptrung gian nay, nen qua 6 loai diem do noi tren co the ve mot dudng dac trUngchung. Tfnh khong doi xufng cua di/dng cong nay the hien d cho: nC/a tren 0 - 300mm cua thanh AR co dp hi#u dung bang 42%, con nCra di/di 300 - 650 mm co dphi#u dung bang 58% dp hi#u dung toan thanh.

- Cung trong tru"dng h0p cap thanh KC1 - KC4 nhung mpt phan 300 - 400mm, doi vdi thanh Cacbua bor, dac tri/ng tfch phan co dang bat doi xufng manh hdn :Phan tren cua thanh Cacbua bor co dp hi#u dung bang 38%, con phan dudi 62%.

c) Dp hieu dung cua thanh AR phu thupc vao cau hinh vung hoat va vdi cungmpt cau hinh, phu thuoc manh vao vi trf cap thanh KC1-KC4. Sau day la torn tat dphieu dung da xac dinh du"dc.

- Ddi vdi thanh Cacbua bor:

+ Cau hinh 75 BNL va 11 thanh Berili, xung quanh thanh AR co 1 BNL, 4thanh Berili, 2 thanh nhom : dp hieu dung = 1,13.

+ Cau hinh 74 BNL va 13 thanh Berili, xung quanh thanh AR co 3 thanh Beriliva 4 thanh nhom : dp hieu dung = 0,86.

41

Hinh 8. Cac d$c trims t£ch phan cua thanh AR bang

thep khong gi (D9 hifu dynfi; cla chuan ve ctcm vl)

Hinh 9. D|c trirng t£ch phan cua thanh AR trong

cau hinh lam vifc cua lo gom 88 BNL.

+ Cau hlnh 86 BNL va 13 thanh Berili, xung quanh thanh AR co 3 thanh Beriliva 4 thanh nhom : dp hi#u dgng = 1.04.

+ Cau hinh 83 BNL va 15 thanh Berili , xung quanh thanh AR co 5 thanh Beriliva 2 thanh nhom : dp hieu dung = 0,87.

Cac gia tr| tren d'eu Idn hdn gidi han tren cho phep cua quy tac an toan hatnhan.

- Ooi vdi thep khong g] trong cau hlnh 88 BNL va 18 thanh Berili, dp hi#udung nam trong khoang 0,4 - 0,61

Vi trfKC1 - KC4

650650650

450450

450 - 472

368 - 392360 - 387358 - 383

cac thanhKC2 - KC3

375 - 395363 - 385363 - 395

450450

450 - 472

650650650

Bo hieu dungthanh AR

0.440.460.38

0.440.470.49

0.590.530.56

Cac ket luanve do hieu dung va* dac trung tfch phan thanh AR

1) Khong dung Cacbua Bor lam thanh dieu khien \\J dpng ddi vdi 16 phan ufngDa Lat vi 6q hi#u dgng qua Idn, quy tac an toan hat nhan khong cho phep.

2) Tuf ngay 17/1/1984 da thiet lap cau hlnh lam vi#c chfnh thtfc cua vung hoatgbm 88 BNL va 18 thanh Berili vdi thanh AR bang thep khong gl co dp hieu dgng 0,4- 0,6 tuy theo v| trf cac thanh dieu khien KC1 - KC4.

3) Btfdng dac tri/ng tfch phan thanh AR co dang ddi xufng khi cap thanh KC1 -KC4 6 vj trf nhung hoac rut hoan toan.

Do do khi sCf dgng thanh AR de do dp hi#u dgng cac thanh KC hoac xac djnhcac hi#u ufng nhi#t dp, cong suat, nhiem dpc Xenon ... can chu y den 2 dac diem noitren va sCf dgng dt/dng dac tri/ng tfch phan thfch hdP-

Hlnh 9 torn tat d&c trUng tfch phan thanh AR trong cau hlnh lam vi#c vdi cac vjtrf khac nhau cua KC1 - KC4 : 650, 450, va 380 mm. Trong hoat dpng sau nay cuaId, 6 trang thai tdi han cac thanh KG thtfdng difdc d|t d'eu nhau, vung 450 mm, do docan su1 dgng dtfdng dac trUng tfch phan thanh AR ti/dng ting vdi trirdng hdp nay. Khido dp hi^u dgng thanh AR la 0.49.

44

V - DAC TRtfNG TfCH PHAN CAC THANH KC VA AZ D6\ VC5l CAU HlNH74 BNL

D&c tri/ng tfch phan cua cac thanh KC di/Qfc xac djnh bang phi/dng phap butrCf nhd thanh AR. Tuy nhien do anh hi/dng cua cap thanh KC1- KC4 nen chl dungJhanh AR de xac djnh d&ctri/ng tfch ph^n cua cac thanh KC2 va KC3, con cac thanh KC1 va KC4 se dUtfc xacdjnh qua cac thanh KC2 va KC3.

Cau hlnh 74 BNL, dung trong ngay 22 - 25/12/1983, co 3 thanh Befili va 2thanh nhom bao quanh thanh AR Cacbua Bor. j

5.1 - Dac trtfng tfch phan cac thanh KC2 v£ KC3

Cau hlnh dat tdi nan vdi vj trf cac thanh dfeu khienKC1 = KC3 = KC4 = AR = 0, KC2 = 442 mm,

ho$c KC1 = KC2 = KC4 = AR = 0, KC3 = 487 mm.Do do ta chl xac djnh diTtfc dac tri/ng tfch phan thanh KC2 trong khoang 0 - 442 mmva KC3 trong khoang 0 - 487 mm.

- Ket qua do ngay 24/12/83 6U$c cho trong Bang 5, d day d|c tri/ng tfch phancua thanh KC2 ch? xac djnh dUtfc trong khoang 0 - 333 mm. Be noi ti£p ta sti dgngcac so li#u do thanh AR ngay 22/12/83 trong do c6 cung cac vj trf KC1 - KC3 = KC4= o mm. Cac so li^u nay d\J0c cho trong phan cuoi cua Bang 5, trong 66 gia trj d$hi#u dung r cua KC2 tai vj trf 317 mm di/Qfc suy ra ttf do thj d&c trUng tfch phan KC2theosoli^u ngay 24/12/1983.

BANG 5 - Bu trtr thanh KC2 va KC3 bang thanh ARCau hlnh 74 BNL 13 thanh Be, KC1=KC4=0.

KC2, mm000193193270270333333317340372405455

KC3, mm487398350350280280202202000000

AR, mm03004502254502254502254536504503252250

dr AR-

0.410

0.3000.457

0.4570.4570.4570.457

0.457-

0.1530.234

0.2230.247

45

Hlnh 10, D^c^trirnR tich phnn cua cac thanh KC va ARd^i 'vo'i c6u hlnh vung hoat 74 BNL.

TCr Bang 5 suy ra dc) hi#u dung cua cac thanh KC2 va KC3 theo vj trf, Bang 6.Hinh 10 tr)nh bay cac d&c tri/ng tfch phan (cac di/dng gach noi dung de ngoai suy dphieu dung tong).

BANG 6 - Dac trUng tfch phan thanh KC2 va KC3.

KC2, mm0

193270333317340372405455

dr AR

0.4570.4570.457

0.1530.2340.2230.247'

r KC20

0.4570.9141.3711.2401.3931.6271.8502.097

KC3, mm487398350280202

0

dr AR

0.4100.3000.4570.4570.457

r KC32.0811.6711.3710.9140.457

5-2- Dac trUna tfch phan cac thanh KC1 va KC4

Ngay 24/12/83 xac. djnh dac tri/ng tfch phan cua thanh KC4 doi vdi v| trf KC1 =KC3 = AR = o mm.

Ngay 25/12/83 xac djnh dac tri/ng tfch phan cua thanh KC3 doi vdi vi trf KC2 =KC4 = AR = 0 mm, Bang 7. Cac difdng dac tri/ng nay cung dupe trinh bay tren Hinh10.

BANG 7 - Dac tri/ng tfch phan thanh KC1 va KC4

KC15355004003002001000

KC30

113202298382460490

dr KC3-

0.180.260.560.560.38-0.08

r KC12.021.841.581.020.460.08

0

KC46505004003002001000

KC275

118217300373422448

dr KC2-

0.100.400.450.530.530.14

r KC41.951.851.451.000.470.14

0

Tu" Hinh 10 ta dupe dp hieu dung cua cac thanh KC nhu sauKC1 : 2.08, KC2 : 2.48, KC3 : 2.36, KC4 : 1,96.

Dp phan tfng du" trff cua cau hinh 74 BNL dUo"c xac dinh qua KC2 la 2.1 (KC2 = 455;KC1=KC3=C4=AR=0) va qua KC3 la 2.08 (KC3 = 487; KC1=KC2=KC4=AR=0) :Dp phan ufng du" trCT trung binh bang 2.09.

47

06 sau dudi tdi han, khi nang ca 2 thanh AZ len khoi vung hoat con cac thanh KC vaAR dtfdi, du"0c tfnh bang hi#u gitfa tong dp hi#u dung cac thanh KC va AR vdi dophan ufng du" truf. Vay dp sau dudi tdi han bang

(2.08 + 2.48 + 2.36 + 1.96 + 0.86) - 2.09 = 7.65.

5.3 - Do hieu dung dono thdi cua cap thanh KC1 va KC2

Bang 8 trinh bay dQ hi#u dung cua cap thanh KC1 va KC2 dU0c xac d|nh quaKC3 vdi v| trf cac thanh khac KC4 = AR = 0 (ngay 24/12/83), cung vdi dp hi#u dungtong cong cua KC1 va KC2 tfnh rieng biet. Hai do thj , Hinh 10, dhd (KC1 va KC2) vadhd (KC1) + dhd (KC2) khong khac nhau trong mien 0 - 300 mm, do do hieu ufnggiao thoa giufa 2 thanh KC1 va KC2 chi/a dUdc xac djnh (nho).

BANG 8 - DQ hieu dung dong thdi cap KC1 va KC2

KC1= KC2mm

2982502001501000

KC3mm

0223307375438487

Dp hi#u dung(KC1vaKC2)

2.081.521.000.560.20

0

(KC1)1.020.700.460.200.08

0

(KC2)1.120.780.500.300.16

0

(KC1)+(KC2)2.141.480.96

' 0.500.24

0

5.4- Do hieu dung cua cac thanh AZ1 va AZ2

Be xac djnh do hi#u dung cua cac thanh AZ1 va AZ2 ta trao ddi vai tro cuachung vdi hai thanh KC2 va KC3. Do dieu ki#n day cap di#n cua dc)ng cd AZ2 trenthu"c te chl trao doi dtfdc AZ1 vcTi KC3 vdi AZ2 vdi KC2. Do do tren bang dfeu khienkhi ta thao tac KC3 (KC2) thi trong vung hoat thi/c chat la djch chuyen thanh AZ1(AZ2). Phi/dng phap dUdc ap dung la xac djnh do hi^u dung cua mQt doan thanh AZrbi ngoai suy do hiqu dung tong bang each su" dung dudng dac trifng tfch phan cuacac thanh KC Ian can.

V| trf KC2 = KC3 = 0, gitf vai tro cua thanh an toan.Vdi trang thai tdi han KC1 = KC4 = AZ2 = AR = 0 va AZ1 = 473, dUa AR

xudng v| trf 650 thl phai bu trCr AZ1 = 337. NhU vay doan 337 - 473 cua AZ1 co dQhi£u dung bang 0.86. Tren doan nay hai thanh KC Ian can KC1 co 6Q hi#u dungbang 0.63 va KC2 co do hi#u dung bang 0,74. Do do do hi#u dung cua AZ1 tfnh ty letheo dudng dac trirng KC1 bang 2.82 va theo dac triTng cua KC2 bang 2.97. Gia trjtrung binh cua do hieu dung thanh AZ1 bang 2.9.

Tifdng tu", vdi trang thai tdi han KC1 = KC4 = AZ1 = AR = 0 va AZ2 = 446,dtfa AR xudng vj trf 650 thl phai bu trCr AZ2 = 328. Nha vay doan 328 - 446 cua AZ2co do hieu dung bang 0.86. Tren doan nay hai thanh KC Ian can KC3 co do hi#u

48

dung bang 0.68 va KC4 co dp hi#u dung bang 0.56. Do do dp hi#u dung cua AZ2tinh ty le theo difdng d$c tri/ng KC3 bang 2.98 va theo d&c tri/ng KC4 bang 3.0. Giatrj trung blnh cua dp hi#u dung thanh A22 bang 2.99.

Tom tat ket qua dac tri/ng cau hlnh 74 BNL cho 6" Bang 18.

VI - DAC TRl/NG TICH PHAN CAC THANH KC DOI VCfl CAU HlNH 86 BNL

Cau hlnh 86 BNL co 3 thanh Berili va 2 thanh nhom bao quanh thanh ARCacbua Bor, cac phep do tien hanh ngay 29/12/83, tifdng tu" nhu" muc V.

6-1 ' Dactrtfng tich phan cac thanh KC2 va KC3

Cau hinh dat tdi han vdi vj trf cac thanh dieu khien KC1 = KC4 = 515 va cacthanh khac di/0c cho trpng Bang 9.

BANG 9 - Bu trCr thanh KC2 va KC3 bang thanh ARCau hlnh 86 BNL 13 thanh Be, KC1=KC4=515

KC2, mm00165165228228282

• 282328328377377427427493545

KC3, mm650457457398398343343292292237237165165000

AR , mm2004002004002004002004002004002004002003782000

dr AR-

0.4540.454

0.454

0.454

0.4540.4540.454

0.4540.4540.4540.4540.4540.3800.3800.196

SCr dung dac tri/ng tfch phan thanh AR do ngay 29/12/83 ti/dng ting vdi cac vjtrf KC1 = KC4 = 360 - 400 mm vdi cac gia trj

dQ hieu dung tong cua AR =1.044,dp hi#u dung AR 200 - 400 mm = 0.454,dp hieu dung AR 200 - 378 mm = 0.380,dp hieu dung AR 0 - 200 mm =0.196,

tu" Bang 9 ta suy ra dp hieu dung cac thanh KC2 va KC3 theo vi trf, Bang 10.

49

Hinh 11. Dfc trurns t{ch phan cua cac thanh KC va

AR doi vo'i cSu hinh vung hoat 86 BNL.

BANG 10 - D&c tri/ng tfch phan thanh KC2 va KC3

KC2, mm0

165228282328377427493545

dr AR-

0.4540.4540.4540.4540.4540.4540.3800.196

r KC20

0.4540.9081.3621.8162.2702.7243.1043.300

KC3, mm0

165237292343398457650

dr AR-

0.3800.4540.4540.4540.4540.4540.454

r KC30

0.3800.8341.2881.7422.1962.6503.104

6.2 - Dac trUna tfch phan cac thanh KC1 va KC4

Xac djnh dac tri/ng tfch phan cua thanh KC4 qua KC2 doi vdi v| trf KC1 = KC3= 515 mm, AR = 200 mm.

Xac djnh dac tri/ng tfch phan cua thanh KC1 qua thanh KC3 doi vdi vi trf KC2= KC4 = 482 mm, AR = 200 mm va qua KC2 doi vdi KC3 =KC4 = 448, AR = 200,Bang 11, cung vdi dac tri/ng tfch phan cua AR trong cau hinh 86 BNL.

BANG 11 - Dac tri/ng tich phan thanh KC1 va KC4

KC1 '6505004003002001000

KC16505534002801520

KC30

123227317400465498

KC2190200300400500650

dr KC3-

0.240.500.740.740.430.14

dr KC20.100.850.950.600.30

-

r KC12.792.552.051.310.570.14

0

r KC12.802.701.850.900.30

0

KC46505004003002001000

KC20

108197282347392407

dr KC2-

0.250.400.720.630.430.13

r KC42.562.311.911.190.560.13

0

Tu" Hinh 11 ta di/0c dp hi#u dung cua cac thanh KC nhtf sau :

51

KC1 :2.80, KC2: 3.48, KC3 : 3.10, KC4 : 2.56.Ta cung nhan thay rang trong ca hai phep do qua KC2 va KC3 do hi#u dune

cua KC1 nhu" nhau nhirng dang cac dudng dac trUng thl khac nhau: qua KC3 cd dancdoi xufng qua vi trf 300 mm con qua KC2 dac trUng tfch phan co dang khong doxufng, nCra phan tren co do hi#u dung be hdn ntfa phan dudi. Dieu nay co thenen do anh htfdng cua thanh KC2 len thanh KC1 vl trong qua trinh do 2 thanhchuyen dong ngi/dc chieu nhau.

6.3 - He so giao thoa va do phan ting du trG

Ngay 29/12/83 do do hi#u dung khi d|ch chuyen dong thdi 4 thanh KC, di/dc :AR 650 500 400 300 200 100 04 KC 363 367 375 385 392 398 400.

TCr do du"dc do hieu dung 4 thanh KC trong khoang 363 - 400 :dhd 4KC (363-400) = dhd AR = 1.04.

Mat khac, xac djnh do hi#u dung tong cua tCrng thanh rieng bi#t neu gia thiet su" dungcac du"dng dac tru"ng tren Hinh 1 1 :

dhd KC1 (363-400) = 2.04 - 1.76 = 0.28dhd KC2 (363-400) = 2.48 - 2.12 = 0.36dhd KC3 (363-400) = 2.22 - 1.90 = 0.32dhd KC4 (363-400) = 1.90 - 1.66 = 0.24

Tong = 1.20Nhu vay h# so giao thoa cua 4 thanh KC bang

K = 1.04/1.20 = 0.87.Do hi#u dung cua tong 4 thanh dong thefi bang do hi#u dung tong cua tCrng thanhnhan vdi h# so giao thoa

dhd 4KC & = k x dhd 4 KC = 0.87 x 11.94 = 10.40.

Ttfdng ufng vdi trang thai tdi han KC1=KC2=KC3=KC4= 385 mm va AR= 300mm,dhdAR (300) = 0.38,dhd KC1 (385) =1.94, dhd KC2 (385) =2.34,dhd KC3 (385) =2.10, dhd KC4 (385) = 1.80, Tong 4 KC « 8.18.

Do phan ufng di/ triFbang 0.87 x 8.18 + 0.38 = 7.50.Do sau dudi tdi han tuong ufng vdi cac thanh KC va AR nam trong vung hoat

bang (10.40 + 1.04)- 7.50 = 3.94.Tom tat cac ket qua doi vdi cau hinh 86 BNL dUtfc cho trong Bang 18.

VII - DAC TRQNG TICH PHAN CUA CAC THANH KC VA AZ o6l V(3l CAUHJNH 88 BNL

Trong cau hinh 88 BNL 18 thanh Berili, thanh AR bang thep khong gi.Ngay 14/1/84 xac d|nh dac trUng tfch phan cac thanh dieu khien nhi/ng do do

phan ufng dU truf qua Idn khong the do dU0c toan bp mpt thanh. Do do ngay 14/1/84(chJ mot ngay) da dat mot thanh Cacbua Bor vao bay ndtron de giam do phan ufng du*

52

trur. Vi#c xac djnh do hi#u dung cac thanh KC va AZ trong dieu ki#n nhu" vay khonghoan toan sat vdi thu"c te.

7.1 - TrUdna hap b)nh thUdna. 66 hiiu dung KC2 va KC3

Trong bay ncftron khdng c6 thanh Cacbua Bor , ngay 14/1/84 da xac djnh d$ctrUng tfch phan c£c thanh KC2 va KC3 qua thanh AR vdi vj trf KC1 = KC4 = 650 mm.Xua't phat tCf vj trf KC2=298, KC3=650, AR=200, tuan ttf dUa AR xuong 400 va bu tru"bang KC3 len 512, roi lai rut AR len 200 va bu trCr bang KC2 xuong 322 cho den vj trfcuoi cung KC2=650, KC3=233, AR=650 (xem them Bang 12) :

KC2 298 298 322 ..( xem .. 545 650 650 ; dhd AR (200-400) = 0.244,KC3 650 512 512 ..Bang.. 263 263 233 ; dhd AR (287-400) = 0.150,AR 200 400 200 .. 12).. 400 287 650 ; dhd AR (287-650) = 0.260.

Vi khong the rut toan bo thanh KC2, (ho|ic KC3), ra kh6i vung hoat, g<?i x =dhd KC2 (0-298), /y = dhd KC3 (0-233)/, sCf dung cac gia tri tren day cua do hi#udung tCrng phan cua thanh AR trong dieu ki?n KC1=KC4=650 (do ngay 23/1/84), tadU0c dac trung tfch phan cac thanh KC2 va KC3, Bang 12.

BANG 12 - Bo hi#u dung KC2 va KC3 qua AR vdi KC1=KC4=650 mm

KC2298322345368 t

395 '425457493545650

KC3650-512512-468468-434434-404404-373373-343343-313313-290290-263263-233

dhd KC2-X0 (+x)0.2440.488.0.7320.976

. 1.2201.4641.7081.9532.102

KC3233263290313343373404434468512650

KC2650

650-545545-493493-457457-425425-395395-368368-346345-322322-298

298

dhd KC3-y0 (+y)0.2600.5040.7480.9921.2361.4801.7241.9682.2122.456

Vdi thanh Cacbua Bor trong bay cho phep xac djnh toan thanh KC3 va bo sungphan c6n lai cua thanh KC2, tu" 66 suy ra gia trj cua x va y (xem sau day).

7.2 - TrUdng hop Bay natron co thanh Cacbua Bor

Ngay 16/1/84 xac dinh dac trifng cac thanh dieu khien :- KC2 va KC3 qua AR vdi KCUKC4=352 mm, Bang 13;- KC1 qua KC3 vdi KC2=KC4=346 mm, AR=300 mm va KC4 qua KC2 vdi

KC1=KC3=327 mm, AR=300 mm, Bang 14.

53

Hinh 12. D£c tnrng tich phan cua cac thanh KC

dSi vo'i cfiu hlnh vung hoat 38 BNL.

BANG 13 - Bp hi#u dung KC2 va KC3 qua AR vdi KC1=KC4=352 mmdhd AR (200-400) = 0.27

KC20

135182226260290318347373400435

KC.3650-504504-456456-420420-383393-348348-320320-289289-252252-209209-157

157-0

dhd KC20

0.270.540.811.081.351.621.892.162.432.70

KC30

157209252289320348383420456504

• 650

KC2435

435-400400-373373-347347-318318-290290-260260-226226-182182-135135-0

0

dhd KC30

0.310.580.851.121.391.661.932.202.472.743.01

Hinh 12 trinh bay cac di/dng dac trtfng KC do trong ca 2 tri/dng hop. Dp hi#udung KC3 = 3.01 trong tri/dng hop bay c6 thanh Cacbua Bor. Xap xl gia trj nay vditrirdng hop khong co thanh Cacbua Bor trong bay ta dUOc y = 0.68. Boi vdi thanhKC2 fan dau do trong khoang 298-650 mm, Ian sau co thanh B4C do trong khoang 0-435 mm, neu ta noi xap xl cac gia trj trong khoang chung 298-435 ta di/Oc x=1.41 vadp hieu dung cua KC2 ta dtfOc dp hieu dung cua KC1 = 3.50 . Dp hi#u dung cuaKC4 = 3.15. Dp hieu dgngtong 4 KC = 13.11.

BANG 14 - Bac trUng tfch phan thanh KC1 va KC4, vdi AR=300,KC2=KC4=346 mm KC1 = KC3 = 327 mm

KC10

100200300400473511

KC3650550440335205

00

dr KC3-

0.110.520.860.970.54

AR=0.22

r KC1 j0

0.110.631.492.463.003.22

KC40

100200300400500650

KC2650548452370282198145

dr KC2-

0.130.490.730.850.660.29

r KC40

0.130.621.352.202.863.15

7.3 - He so aiao thoa va do phan ufng dU XrQ

55

Ngay 14/1/84 xac dinh dp hi#u dgng khi dich chuyen dong thdi 4 thanh KC,AR 650 400 200 04 KC 450 457 467 472

Do hi#u dung dong thdi 4 thanh KC trong khoang 450 - 472:dhd 4KC (450-472) = dhd AR =0.49.

Mat khac, xac djnh dp hieu dgng tong cua tifng thanh rieng b>i#t neu su" dgngcac difdng dac trifng tren Hlnh 12 :

dhd KC1 (45(3-472) =3.00-2.82 = 0.18dhd KC2 (450-472) =2.90-2.78 = 0.12dhd KC3 (450-472) =2.50-2.40 = 0.10dhd KC4 (450-472) =2.68-2.52 = 0.15

dhd tong = 0.55

He so giao thoa k = 0.49 / 0.55 = 0.89.

Dp hi#u dgng cua cac thanh KC va AR d vj trf tdi nan ladhd AR (400) = 0.33dhd KC1 (457) = 2.87, dhd KC2 (457) = 2.81 .dhd KC3 (457) = 2.45 dhd KC4 (457) = 2.57

Tong 4KC =10.70Do phan tfng dg1 trCT =0.89 x 10.70 + 0.33 = 9.85

Do sau dudi tdi han ti/dng ufng vdi cac thanh KC va AR nam trong vung hoatbang (0.89x13,11 +0,49) -9,85 = 2,31

-4 - Do hieu dung cua cac thanh AZ

PhUdng phap do nhu" dUOc trinh bay 61 muc 5.4. KC2 = KC3 = 0, giur vai trocua cac thanh an toan.

Vdi trang thai tdi han KC1 = KC4 a AZ2 = 457, AR = 300 va AZ1 = 650,dira 4 thanh dau xuong vi trf 650 thi phai bu trCr AZ1 = 262. Nha v|y doan 262 -650cua AZ1 co dp hi#u dgng bang 2.10. Tren doan nay hai thanh KC Ian can KC1 co dphieu dgng bang 2.48 va KC2 co dp hieu dgng bang 2.45. Do do dp hieu dgng cuaAZ1 tfnh ty le theo dp hi#u dgng 3.55 cua KC1 thl bang 3.0 va thep dp hi#u dgng3.45 cua KC2 thi bang 2.96. Gia tr| trung blnh cua dp hi^u dgng thanh AZ1 bang2.98.

Tirdng tg", vdi trang thai tdi han KC1 = KC4 = AZ1 = 489, AR = 300 va AZ2 =650, di/a 4 thanh dau xup'ng vi trf 650 th) phai bu trCr AZ2 = 320. Nhir vay dpan 320 -650 cua AZ2 co dp hi#u dgng bang 1.42. Tren dpan nay hai thanh KC Ian can KC3co dp hieu dgng bang 1.60 va KC4 co dp hi#u dgng bang 1.64. Do 66 dp hi#u dgngcua AZ2 tfnh ty le theo dp hi^u dgng 3.01 cua KC3 thi bang 2.67 va theo dp hieu

56

dgng 3.15 cua KC4 thl bang 2.73. Gia tr| trung blnh cua dp hi#u dgng thanh AZ2bang 2.70.

Tom tat cac ket qua do doi vdi cau hlnh 88 BNL diftfc cho trong Bang 18.

VIII - 0 0 HIEU DMNG CUA MOT SO YEU T6 TRONG VUNG HOAT

Do dp hi#u dgng cua mpt so bo nhi£n li#u, thanh Berili va cac kenh thfnghi#m trong vung hoat doi vdi cac cau hlnh khac nhau. Dung phtfcfng phap bu trCfde so sanh vdi dp hieu dgng cua cac thanh KC va AR.

8-1 " £ o hieu dung cac bo nhien lieu

Dp hieu dgng cua mpt so bo nhien li#u d cac vj trf khac nhau diipc xac d|nhbang each du"a 16 phan ufng len trang thai tdi nan khi co BNL va khi khong co BNL, 6mang de trdng (co nu"dc) ho^c du"pc thay bang thanh nhom chen hay thanh Berili .Ket qua cho trong Bang 15.

BANG 15- Dp hieu dgng cua BNL(cot 1 : so BNL, vj trf, chat thay the, dp hi#u dgng,(cot 2 : dich chuyen thanh va dp hi#u dgng tUdng ufng)

BNL, vj trf/dp hieu dgng

-N 067,. 10-1

ni/dc 0.58- N 060, 7-1

ni/dc 0.40- N 188, 7-4

nude 0.36- N 186, 6-3

nu*dc 0.19- N 194, 7-5

nirdc 0.16- N 067,11-

1Al 0.52

-N 067, 11-1Be 0.28

KC1

a)

213-2130

213-1720.18

213-1720.18

172-1720

172-1720

213-2130.

213-2130

Dich chuyen khi khong co BNL/ dp hieu dgncKC2

Cau hlnh 74

213-213

213-1720.17

213-2720.17

172-1720

172- 1720

213-2130

213-2130

KC3

BNL (23/12/8:

213-213

213- 1720.18

213 -2720.18

172- 1720

172- 1720

213-2130

213-2130

KC4

3)

213-213

213- 1720.18

213 - 2720.18

172 - 1720.

172- 1720.

270-2130.26

270-2130.28

)AR

213-2130

213-1720.18

213-1720.18

172-1720

172-1720

213-213 '0.

213-2130

57

BANG 15 - (tiep)BNL.vi trf/

dp hieu dung

-N 060, 7-1nude 0.52- N 056, 1-4Al 0.44- N 076, 1-3Al 0.62- Ni58, 1-2Al 0.60-N 076,1-3 &

Ni 58, 1-o

-Al 1.23

-N150, 7-4nUdc 0.48

-N194, 7-3nude 0.16

-N 069, 10-1nUdc 0.28-N186, 6-3nude 0.16

Dich chuyen khiKC1 KC2

b) Cau hinh 86

385-3380.36

420-4200.

385-3850.

390-3900.

385-385

0.

358 - 3580.44

420 - 4200.

385 - 3850.

390 - 3900.

385 - 385

0.

c) Cau hlnh 88

470-4080.48

470-4480.16

470-4320.28

470-4480.16

451-4510.

451-4510.

451-4510.

451-4510.

khong co BNLKC3

BNL (31/12/8

358 - 3380.40

420 - 4200.

385 - 3850.

390 - 3900.

385 - 385

0.

BNL (18/1/84

451-4510.

.451-4510.

451-4510.

451-4510.

/ dp hieu dgngKC4

3)

358 - 3380.36

420 - 4200.

385 - 3850.

390 - 3900.

417-385

0.19

•

451-4510.

451-4510.

451-4510.

451-4510.

AR

0 -650-1.04330-00.40

395-00.62

650-3200.60

650-0\

1.04

300-3000.

300-3000.

300-3000.

300-3000.

^•2 - Do hieu duna cua cac thanh Berili va cac veu to khacDp hi#u dung cua cac thanh Berili dUdc xac djnh ddi vdi nUdc ho^c thanh

nhom chen thay the, dp hieu dung cua cac yeu to khac di/Oc xac djnh doi vdi nude,Bang 16.

Nhin chung, dp hieu dung cua cac BNL va cac thanh Berili deu di/Cng va namtrong khoang 0.1 - 0.6 ngoai tru" khoi Be cua bay noiron, con dp hi#u dung cua cacthanh kim loai (thep khong gl, nhom) doi vdi nUdc la am.

BANG 1 6 - 0 6 hieu dgng thanh Berili va cac yeu to khacYeu to : KTN : kenh thf nghi#m; AL1 : th6i nhom dai 85 cm di/dng kmh 15

mm; AL2 : thanh nhom dai 53 cm treo dudi thanh di'eu khien; Fe : thanh thep khonggl dudng kfnh 29 mm; BeK : Khdi Berili cua bay ndtron

Vi trf: BN : bay ndtron; MQ : mam quay.

58

Yeu to, vj trf/dp hi#u dung

-1 Be, 9 - 1nirdc 0.34

-2Be,9-1 va10-1

ni/dc 0.62-3 Be, 9 - 110-1 va11-1nirdc 0.86

- KTN, 7 - 3nude - 0.12

-KTN BNnirdc 0.08

-A11 BNnirdc 0.04

-A 11 MQnirdc 0.01

-A 12 7 - 3nirdc -0.32

- 1 Be 1-3A 1 0.26

- 1 Be 1-2A 1 0.27

- 1 Be 1-4A 1 • 0.27

- 1 Fe 7 - 1ntfdc -0.58

- 1 Be 13-2ni/dc 0.20

- 1 Be 1-4nirdc 0.26

-A 12 6 - 3nirdc -0.32

- BeK BNnirdc 2.64

Djch <KC1a) Cau

213-2130.

213-2130.

213-213

0.

172- 1720.

172 - 1720.

172-1720.

172- 1720.

172 -1720.b) Cau

385 - 3850.

395 - 3950.

405 - 4050.

338 - 3380.c) Cau

470 - 4400.20

470 - 4340.26

407 - 448-0.32

470- 1472.64

:huyen khi khong cd yeu to / dp hi#u dungKC2 KC3

hinh 74 BNL (25/12/83)213-213

0.213-213

0.213-213

0.

172- 1720.

172- 1720.

172-1720.

172-1720.

172-1720.

hinh 86 BNL385 - 385

0.395 - 395

0.405 - 405

0.338 - 338

0.hinh 88 BNL

451 - 4510.

451 - 4510.

451 - 4510.

451 - 4510.

213-2130.

213-2130.

213-213

0.

172- 1720.

172 - 1720.

172-1720.

172-1720.

172-1720.

. (31/12/83)385 - 385

0.395 - 395

0.405 - 405

0.338 - 338

0.(19/01/84)451 - 451

0.451 - 451

0.451 - 451

0.451 - 451

0.

KC4

213-2130.

213-2130.

213-213

0.

172- 1720.

172 - 1720.

172- 1720.

172- 1720.

172- 1720.

385 - 3850.

395 - 3950.

405 - 4050.

338 - 3380.

451 - 4510.

451 - 4510.

451 - 4510.

451 - 4510.

AR

272-00.34

414-00.62

650-0

0.86

395 - 450-0.12

433 - 3950.08

450 - 4330.04

437 - 4330.01

277 - 438-0.32

247-00.26

232-00.27

232-00.27

335 - 650-0.58

300 - 3000.

300 - 3000.

300 - 3000.

300 - 3000.

IX - PHAN BO TtfONG DOI CUA MAT B$ THONG LU0NG N0TRON NHIET

Xac dinh phan bo mat dp thong lu"0ng noiron nhiet theo tiet di#n ngang vatheo chieu cao cua vung hoat doi vdi caiu hinh co bay ncftron bang phi/cfng phap kich

59

host la do Dy-164. Dong vj Dy-165 co thdi gian 140 phut va phan ra beta vdi nangItfOng Emax = 1.28 Mev (84%) va 1.19 MeV (14%)

9.1 - Cau hinh 86 BNL

Do tai 24 diem, 23 diem d BNL va mpt diem d bay notron. La do d&t tai diemgiufa BNL theo chi'eu cao va 2 diem tren va dtfdi each diem gitfa 5 cm. Ket qua du"dclay trung binh cua 3 diem do. Chieu cac la do d cong suat co 1.E-2% N, tufc la vaichuc Watt, thdi gian chieu la 10 phut. Bang 17 va Hlnh 13; cho phan bo tLfdng doicua mat dp thong Itfdng notron, lay mat dp thong lifdng notron tai tarn bay Bang 1.Hlnh 14 trinh bay phan bo ttfdng doi cua mat dp thong Itfdng notron nhi#t theo tietdien ngang ben trong bo nhien li#u tai cac 6 6-5 va 6 6-8.

Ta thay rang mat dp thong Itfdng notron tang tif ngoai vao trong va nhan giatri cd 0,95 ngay tai mep BNL tiep xuc vdi thanh Berili cua bay notron. TO cac ket quado ta diidc gia tri mat do thong lu"dng trung blnh la 0.54. Do dodQ bat dong d'eu theoban kfnh la

Kr = (|)max/ ((}>) = 0.95/0.54 = 1.76

9.2 - Cau hinh 88 BNL

Do tai 28 diem trong do co kenh i/dt K7-1 va bay ndtron. Bang 17-va Hinh 15cho ket qua do. Mat do thong li/dng trung blnh tren 27 diem do la 0.35. Dp bat dongd'eu khong xac dinh dUdc vi khong do difdc gia tri cu"c dai. Nhom chuyen gia Lien X6xac djnh Kr = 1.84 doi vdi ca hai cau hlnh (xem tai lieu 121).

Nhln chung, qua Bang 17 ta nhan thay rang phan bo mat dp thong liTdng trong2 cau hinh kha khac nhau, mat dp thong Itfdng trong cau hlnh 86 BNL 13 thanh Bephan bd d'eu hdn so vdi cau hinh 88 BNL 18 thanh Be d. day mat dp tai bay ndtrondUdc chuan bang 1.

BANG 17 - Mat dp thong Itfdng ndtron theo tiet di#n ngang(chuan tai bay ndtron ve ddn vj).

Vi trfc.h 86 BNLc.h. 88 BNL

Vj tric. h. 86 BNLc. h. 88 BNL

Vi trfc. h. 86 BNLc. h. 88 BNL

Vi tric. h. 86 BNLc. h. 88 BNL

2 - 20.520.314 - 20.460.255 - 8

0.29

7 - 20.510.32

2 - 40.550.354 - 40.500.27

5- 100.46

7 - 30.570.34

2 - 50.570.344 - 70.490.28

5 - 1 10.470.28

7 - 40.640.36

2 - 70.450.284 - 90.390.256 - 3

0.30

BayN1.001.00

3 - 10.480.315- 10.550.316 - 50.600.40

7 - 80.600.37

3 - 4

0.325 - 30.50

6 - 80.600.38

7 - 90.550.31

3 - 6

0.325 - 4

0.306 - 1 0

0.28

7 - 90.550.31

3 - 90.430.285 - 60.60

K7- 10.520.65

60

•m^-J*Ky.xt'Jm>-

Hinh 13 . Ph^n bo tircng doi cua mat cto thong liromg

no'tron nhiet theo tiet difn ngang vung

ho at vo'i cau hinh 86 BNL.

1.0,...,0.52,...

Bo nhien

Cac thanh bao v§ a\x co AZ1 , AZ2

Cac thanh bu trir KC1 ,..., KC4

Thanh dieu khien tir dong AR

Thanh chen Berili

Thanh chen hhom

Bay no'tron

Phan t>6 tiroTag d5i cua mat do

thong lirqmg no'tron nhif t

61

BNL 6.5 BNL 6.8

\ \\ \ \ N' \ \ \\ \ \ \ V \

0. fr

\ \

Hinh 14 . Phan bo tircng doi cua m|t d§ thong luringno'tron nhiet theo tie't dien ngang cac BNLtai 6 6.5 va 6 6.8 ( cau hinh 86 BNL ).

62

\

kenh

thi nF,hiem :

' Hxnh 151. Phan bo tiro*ng doi cua m§t d$ thong lirqug

no'tron nhift theo tiet difn ngang vung

ho at vo'i cgu hlnh 88 BNL.

Cac BNL

Cac thanh bao vf s\f co AZ1 , AZ2

Cac thanh bu trur KC1,..., KC4

Thanh dieu khien tir d§ng AR

Thanh chen Berili

Bay no'tron

1 , . . . Phan b6 tirong doi cua mat do

thong liroTig no'tron nhi§t

63

i -H

;

C.V i

0., Jtren

•fOD .120 300 . . 400VOL

dirci

6on

Hinh 16. Phan b6 tirong &5± cua m|t d$ thong lirqug no'tronnhift theo chieu cao vung hoat ta i bay no'tron.

64

Hinh 1 7 . phan b5 tircmg d6i m§t d§ thong lirq-ng no ftr6n n h i f ttheo chieu cao vung hoqit, tg.i cac m^t cua 6 5«6

65

Hinh 18. Phnn bo t\wnr, dot mat dq thon^ lirq-ng no'tron nhifttheo tryc nac kenh ngang thi nghifm.

Hmh 19', Phan b6 tuoiig doi mat do thong liroTig no'tron nhift

dQc theo true c§t nhiet.

Trong cau hinh nay cung da do phan bo tifdng doi cua mat dp thong Itfdngndtron nhiet theo chieu cao vung hoat tai bay ndtron, Hinh 16, va tai cac mat BNL vjtri 5-6, Hinh 17..Trong cac phep do nay, v| trf cac thanh KC la 457 mm va AR la 300mm. Ta thay rang phan bd trong bay ndtron khong ddi xufng : dtfdng cong dat cu"c daitai khoang 300-450 mm. Cac phan bd theo chieu cao tai 6 5-6 khong co dang nhu" taibay ndtron. Dp khong dong d'eu tai mat 1 va mat 6 la 1.13; d mat 2 va mat 5 la 1.23;va d mat 3 va mat 4 la 1.29.

Hinh 18 trinh bay phan bd ti/dng ddi mat dp thong li/dng ndtron nhi#t theo truecac kenh ngang thi nghi#m : Mat dp nay giam theo ham mu. TCf bie"n vanh phan xaden mat ngoai be tong bao v# mat dp thong Itfdng giam khoang 500 fan. Kenh ngangso 4 dat sat vung hoat, mat dp thong liiqmg ndtron theo kenh nay giam 1000 Ian tu1

trong ra ngoai. Ngoai ra cung can chu y rang mat dp thong \UQng ndtron d kenh so 1Idn hdn d kenh so 2 : dieu nay khang dinh rang kenh so 1 co phan dan dong rongnam trong vanh phan xa, chuf khong phai la kenh so 2 nhtf da cho trong cac ban vethiet ke cua 16 Triga (cac hinh 3-7).

Hinh 19 trinh bay phan bd tu*dng ddi mat dp thong Itfdng ndtron nhiet dpc theotrue cot nhiet. Mat dp thong Itfdng notron nhiet giam 2 bac tu" trong ra ngoai.

Trong thdi gian khdi dpng vat ly cung da xac dinh mat dp thong Itfdng tuy#tddi cua ndtron nhiet va ndtron nhanh tai mpt so vi trf. Cac ket qua nay se dUdc trinhbay trong phan 3 ve khoi dpng nang li/Ong 16 phan ufng.

X - NHAN XET - KET LUAN

10.1. Tom tat ve khdi dong vat Iv

Muc tieu cua khdi dpng vat ly la nap nhien lieu va xac djnh cau hinh lam vi#c,khao sat cac dac trung vat ly cua cau hinh do.

Cong viec nap nhien li#u bat d'au luc 14h30 ngay 30/10/83. Trang thai tdivdi cau hinh vung hoat khong cd bay ndtron d£t luc 19h50 ngay 1/11/83. Cau hinhgdm 69 BNL vdi khdi lUdng U-235 tdng cpng la 2781.2 g.

Trang thai tdi han vdi cau hinh vung hoat cd bay ndtron dat luc 17h48 ngay18/12/83. Cau hinh gdm 72 BNL, khdi li/p"ng U-235 tdng cpng la 2987.4 g.

Ngay 21/12/83 bat dau nap them nhien lieu de xac djnh cau hinh lam viec vakhao sat cac dac trUng vat ly. Cau hinh dau tien cd dp phan ufng du" trtf nho gom 74BNL, 2977.9 g U-235 . Cau hinh thuf 2 gdm 86 BNL vdi 3461.8 g U-235. Cau hinhlam viec chinh thufc gdm 88 BNL dUp"c thiet lap ngay 14/1/84 vdi khdi lupng U-235tong cpng la 3537 g.

Cac ket qua do dac vat ly difdc KST Ngd Quang Huy xCf ly song song vdichuyen gia N.V. Arhanghenski phu trach nhdm khdi dpng 16 Lien X6 121. Cac ket quavat ly dUOc torn tat trong Bang 18, trong do viet trong ngo&c la cac ket qua xti ly phia

68

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Hinh 20 . D^c trorng tich phan cua cac thanh KC va AR

trong cau hinh 86 BNL ( Arhanghenski ).

69

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0 5 0 0 , ( • 6 0 0 i \_[ jam. i : •; i; i i L I iJ.I i

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Hinh 21 . D|c tru*ng tich phan cua cac thanh KC vet ARtrong cau hlnh 88 BNL ( Arhanghenski ).

70

Lien X6 121 y sai khac nhau 10% trong dai da so trtfdng h0p. Hinh 20 va 21 trinh baycac dac triing tich phan cua cac thanh dieu khien theo tai lieu 121. Sau d&y la ffiQt sonhan xet v'e dac trtfng cac thanh dieu khien trong 3 cau Hinh 74, 86 va 88 BNL.

10.2' Ve do hieu dung

1) DQ hieu dung cac thanh tang theo so Iif0ng cac bo nhien li#u dtf0c nap.Bieu nay rat ro doi vdi 4 thanh KC, ft ro hdn doi vdi 2 thanh AZ, cd the do each do vax(i ly xap xl di/0c dung. Do hi#u dung AZ1 va AZ2 gan nhtf khong thay doi trong 2cau hinh 74 va 88 BNL. Dp hi#u dung AZ nh6 hdn dp hi#u dung KC trong cau hinh88 BNL nhi/ng Idn hdn trong cau hinh 74 BNL . Dp hi#u dung cac BNL thl gan nhu"khong doi vdi 3 cau hinh.

2) DQ hieu dung cac thanh dieu khien khong deu nhau tuy rang ca 6 thanh KCva AZ deu cd cung cau true, va khong tuan theo quy luat doi xufng cua vung hoat(phan nCfa vung hoat phfa AZ1 cd dp hi^u dung cao hdn). Cd the do each tien hanhphep do khdng thfch h0p. Can tim hieu khao sat nguyen nhan hi$n tU0ng nay.

BANG 18 - Tom tat cac thong so vat ly khdi dpng Id

Thong so

- Khoi ludna taiSo va khoi lU0ng U 235

So va khd'i thanh Be

74 BNL 13 Be

74- 2977,9 g13-28211 g

- Do hieu duna cac thanh di'eu khienAZ1 .AZ2KC 1KC2KC3KC4ARTong 4 KChe so giao thoa 4 KCdo phan ufng du* tru1

dp sau dtfdi tdi han

- Do hieu dung BNLd 1-2 so vdi ALd 1-3 so vdi ALd 1-4 so vdi ALd 1-2 va 1-3 -nt-

d 1-2 va 1-3 va 1-4 -nt-

2.90 (2.55)2.99 (2.40)2.08 (2.14)2.48 (2.40)2.36 (2.18)1.96(1.99)0.86 (0.86)

8.881

2.09 (2.09)7.65 (7.48)

86 BNL 13Be

86-3461.8 g13-28211 g

--

2.80 (2.68)3.48 (3.40)3.10(3.04)2.56 (2.48)1.04(1.04)10.4(1.04)

0.877.50(8.14)3.94 (3.50)

0.600.62 (0.54)

0.441.22 (1.03)1.66 (1.42)

88 BNL 18Be

88-3537 g18-33113 g

2.98 (2.60)2.70 (2.39)3.50 (3.28)3.45 (3.45)3.01 (3.00)3.15 (2.74)0.49 (0.49)11.67 (11.0)

0.899.85(9.17)2.31 (2.33)

71

BANG 18- (tiep)

Thong so

6 6 - 36 7 -16 7 -36 7 - 46 11-16 11 - 1 so vdi AL6 11 - 1 so vdi Be- Dp hi#u dung cac thanh

Be 6 1-2 so vdi ALBe 6 1-3 - nt -Be 6 1-4 - nt -Be 6 1-4Be 6 1-2 va 1-3Be 6 1-2 va 1-3 va 1-4Be 6 9-1Be 6 9-1 va 10-1Be 6 9-1 va 10-1 va 11-1Be 6 13-2Khoi Be bay ndtronKenh thf nghiem 6 7-3Kenhthf nghiembayndtronTh.AL 85mm bay notronTh.AL85mm mam quayTh.AL 530mm 6 6-3Th.AL 530mm 6 7-3Th.thep khong gf 7-1

74 BNL

0.100.400.160.380.580.520.28

Be va

0.340.620.88

-0.120.080.040.01

-0.32

13 Be

(0.19)(0.60)(0-16)(0.55)(0.58)(0.47)(0.16)yeu to

(0.30)(0.62)(0.93)

(-0.12)(0.08)(0.04)(0.01)

(-0.37)

86 BNL 13Be

0.52 (0.53)

khac0.270.260.27

0.53 (0.42)0.80(0.61)

-0.58 (-0.46)

88 BNL 18Be0.16(0.18)

0.16(0.18)0.48 (0.52)

0.26 (0.30)

•0.20 (0.25)2.64 (2.44)

-0.32 (-0.34)

3) He so giao thoa giam khi do hi#u dung cac thanh tang. Dieu nay the hiend tfnh chat bat doi xOfng cua cac dUdng d|c trtfng tfch phan khi cac thanh nhung mQtphan trong vung hoat. Dieu nay cung con the hi$n khi thanh AR bang Cacbua Borlam cho do hi#u dung cua cac thanh Ian can KC1 va KC4 nho hdn dO hi#u dgng cuaKC2 va KC3. Do vai tro quan trQng cua h$ so giao thoa trong vi#c xac d|nh do saududi tdi han, Can khao sat chfnh xac h# so' giao thoa trong cau hinh lam viec cua cacthanh dieu khien.

10.3 - Ve dang cac dac trUng tfch phan

Trong ca 3 cau hinh, dac trUng tfch phan cua cac thanh KC2 va KC3 du"dc xacdinh bang each bu trCf so sanh vdi thanh AR, con cac thanh KC1 va KC4 di/dc xacdinh qua KC2 va KC3.

72

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74

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Hinh 24. D § C trung tfch phan cua thanh KC4 vaid§ hifu dung chuan ve den vi

75

- Dac trtfng tfch phan thanh KC2 khong dirtfc xac dinh cho toan thanh trong ca3 cau hinh.

- Doi vdi KC3 dac trting toan thanh chl dtfdc xac dinh trong cac cau hinh 86BNL (trong cau hinh 88 BNL phai dat them thanh Cacbua Bor trong bay ndtron).Hinh 22 trinh bay cac dac trtfng tfch phan nay (da dtfdc chuan v'e ddn vi) : ca 2di/dng dac trtfng nay d'eu khong doi xufng qua v| tri trung tarn 300 mm. Nguyen nhanco the do cac thanh KC1 va KC4 nhung lung chiTng vung hoat (tfnh bat ddi xting ronet hdn vdi KC1=KC4=352 mm trong cau hinh 88 BNL).

- Dac trUng tfch phan toan thanh KC1 trong cau hinh 86 BNL dtfdc xac djnh 2Ian khac nhau, dung KC2 hoac KC3. Cac dieu kien cau hinh cac thanh di'eu khientrong 2 tri/d'ng hop ttfdng tg" nhu" nhau. Cac di/dng dac trtfng ti'ch phan chuan, Hinh23 , neu ro anh hudng cua cac thanh KC2, KC3.

- Dac trtfng tfch phan KC4 dUdc xac djnh trong ca 3 cau hinh va d'eu dungKC2, nam doi xOfng qua tarn, each xa KC4. Anh htfdng cua vi tri cac thanh con lai doivdi cac dirdng dac tri/ng chuan nay dUOc minh hoa tren Hinh 24

Qua phan ti'ch neu tren ta thay ro hieu ufng giao thoa giufa cac thanh. Hieuufng giao thoa (thu"c nghiem cho thay la giao thoa am nghTa la dQ hieu dung dong thdinho hdn tong do hieu dgng tt/ng thanh), khong phat hien dUOc trong cau hinh 74BNL (co do phan ufng di/ truf nho) tang len khi tang li/dng nhien li#u nap vao, nghTala tang do hieu dgng cua moi thanh. Hieu ufng giao thoa nay khong the loai trCr trongcau hinh lam viec ma cac thanh dieu khien nhung deu nhau.

XI. PHU LUC

Nhien lieu Uran dtfdc bao v£ trong vo boc bang nhom day 0,9 mm. Dung vemat an toan hat nhan, lam hu" hai den Idp vo bQC nhien lieu la mot sg* co nghiemtrong.

11.1. Hien tuana auan sat

Ngay 2/11/83, sau khi dat tdi han vdi cau hinh vung hoat khong co bay ndtron,da du"a Id v'e trang thai sau du"di tdi han va du"a cac BNL so 056, 058, 059. 060 va061 \(i 6 6 trung tarn vao cac coc chute tarn 1-1 va 1-2 d trong be 16. Ngay 5/11/83 laycac BNL nay ra xem thi thay chung bi xam. NhUng sau do xem cac BNL con lai trongvung hoat thi thay khong bj xam. Tuy nhien cung da quyet djnh rut tat ca cac bo BNLtu" vung hoat ra, dat lai vao kho nhien li#u, ( Phong 125) va dat cac thanh nhom chenvao vung hoat. Nhtfng ngay tiep theo hoan chfnh cac h£ cong nghe 16, xac dinh chatluong nude va loc nirdc be \6 de loai tru" hi#n tUOng tren.

76

Ngay 14/12/83, chi/a tim dUdc nguyen nhan hien ttfdng xam nhien li#u, va vivay chu"a co bi#n phap xu" ly. Tuy nhi£n de tiep tuc chi/dng trinh khdi dpng 16, van batdau nap nhien li#u trd lai va dat yeu cau theo doi thudng xuyen hi#n tu"0ng nay. Ngay17/12/83 nap nhien lieu vao vung hoat vdi cau hinh co bay natron va dat trang thaitdi han luc 17h48 ngay horn sau.

Ngay 18/12/83 luc 10h30 khi rut thanh nhom chen 6 11-8 da rut nham BNL086 d o 11-7 va rat ngac nhien nhan thay BNL bj xam nham nhd tuy mdi dude datvao vung hoat co 19 gicf. Sau do luc 13h30 rut BNL d 6 6-3 ra xem, cung thay bixam, thanh nay nam trong vung hoat dUOc 21 gid.

Cac ngay 19 va 20/12/83 dC/ng cong viec khdi d<?ng de xac djnh nguyen nhanxam vo thanh nhien lieu. Rut them mpt so BNL ra xem, thay bj xam ca. Dae biet BNL186 d 6 6-3 hien tJOng xam khac han ngay 18/12 : xem luc 14h30 ngay 19/12 cacmat bi xam deu chur khong xam nham nhd ntfa. Nhan xet chung la qua trinh xam xayra kha nhanh, luc dau xam nham nhd den bong va nhu" dUOc ho" tren ngon den dauhoa, Sau do xam md, khong bong va xam dku. NhQng ngay tiep theo tim kiemng;uyen nhan cua hi#n tu"dng d cac bpt khf hoac nguyen nhan di#n phan trong nifdc16.

71.2. Tim hieu nguven nhan

a) Di/ doan dau tien cua hi^n tU0ng xam cac BNL trong cac coc chtia tarn 1-1va 1-2 la an mon Clo. Do do da phan tfch nude trong 2 coc nay va thay ham lifdngClo dat tdi 180 micro g/1 trong khi Clo trong vung hoat khong dU0c viroi qua 50 microg / 1 . Nhu" vay, cho rang da tim dung nguyen nhan va de khac phgc can phai cho hoatdpng he thong Ipc ntrdc vong 1.

He thong nadc vong 1 da lam viec tCr ngay 3/11/83 va sau do van tiep tuc chochay, dong thdi cho he thong loc nude 16 hoat dpng. Tuy nhien sau do hien tUOngxam van con, xuat hi#n ca trong vung hoat. Cac thanh nhom chen bj xam dam 6 dudiva md dan len tren. Dat thtr mot thanh nhom chen vao 6 trung tarn, 2 ngay sau laylen thay bi xam loang \6.

Lay ni/dc \ti vung hoat len phan tfch thi thay ham li/p'ng Clo du"di miJc 50 microg / I . Nhu" vay hien tu"0ng xam co the khong lien quan den si/ tang Clo. De khangdinh dieu nay da lay nu"dc trong 16 va them Clo vao den nbng dp 5 mg /1 (100 Ian Idnhdn mure cho phep) va dat vao do mieng nhom cung thanh phan vat lieu vdi thanhnhom : ket qua khong thay bi xam.

Nhu" vay, hi#n tU0ng xam vo BNL hay vo thanh nhom chen khong phai do anmon Clo ma do nguyen nhan khac.

b) Nhan xet rang ngay 3/11 khi cho nude vong 1 lam viec thay co rat nhieuchat ban, co the day la nguyen nhan. Ngay 12/11/83 ru"a thung 16 va thay nude mdi:

77

trade net hut phan nu"dc sach d tren, cho den gan vung hoat, do vao be chufa nhienlieu da chay; sau do dung bdm xjt khuay nu"dc trong 16 va rifa sach cac b<? phantrong thung 16, ni/dc ban cho qua bo loc co va loc trao doi ion cu cua 16 Triga; sau khirCra sach nhu" vay lai hut nu"dc tCr be chtifa nhien li#u vao thung 16.

Tuy nhien sau khi ru"a sach thung 16, hi#n ti/Ong xam be mat cac thanh chentrong vung hoat van tiep tuc Luc nay ca cac bQ phan khac bang nhom, nha daygieng treo, mam quay .v.v... cung bi xam den. Nguyen nhan van chtfa tim difdc.

c) Ngay 19/12/83 nhan thay rang trong vung hoat co bQt khi noi len kha nhieu,15-20 bQt trong 1 phut. TO 66 co the neu len 2 gia thiet ve nguyen nhan xam be matcac BNL do oxy hoa :

- BQt khi tCr ngoai vao vung hoat qua he thong nude vong 1;- BQt khi' xuat hien do hien tu"0ng dien phan.

Ngu6n goc thuf 2 dUdc khao sat ngay, tim xem nguon dien nao co the gaynen hien ttfdng dien phan cue bQ. Nguon nay khong can cao the nhtfng cong suatphai du Idn, co the la do su1 chenh lech the giufa dat vat ly va diem trung hoa. Do daccho thay la 2 he dat nay dUdc noi vdi nhau trong be 16.

Dong thdi da tien hanh do hieu dien the dung 2 cay sao bang nhom noi vdiampe-ke va phat hien di/dc dong cd 100 micro A khi 2 cay sao de each nhau vanhung vao ni/dc 16. Lam thf nghi#m di#n phan trong binh thuy tinh vdi 2 di#n cu"cbang nhom co cung vat lieu vdi vo BNL sau 2 gid cung thay co hien tUdng xam denva co bQt khi. Tuy nhien trong thf nghiem nay thay trong nude co rat nhieu nhomtrong khi do nbng do nhom trong nu"dc 16 van d mufc 15 micro g / 1 . .

Nhtf vay hien ti/dng xam be mat BNL ft co kha nang do dien phan gay ra. Conlai nguyen nhan oxy hoa do bQt khong khf tu" he nude vong 1 dtfa vao.

TrUdng hdp nay co the xay ra trong nhufng ngay dau khi cho he nu*dc vong 1chay, vdi lu"u lu"0ng nirdc rat Idn va may do li/u lu*0ng nude lai chu"a hoat d<?ng dUOc :luc do chi quan sat thay khong khi tCr ngoai dUdc hut vao cac ong cua he do vi sai .Ngay 21/12/ dat 3 BNL so 087, 180 va 187 vao cac 6 12-2, 12-3 va 12-4 de tiep tucquan sat. Ngay 22/12 rut ra xem khong thay bi xam va trong vung hoat cung khongthay bQt khi noi len. Ngay 25/12 xem lai, thay cac BNL 180 va 187 sang binh thudng,con BNL 078 chi bi xam md d 2 mat. Trong cac ngay nay thlnh thoang cho h# ni/dcvong 1 va vong 2 hoat dong. Den ngay 28/12 xem lai cac BNL thay thanh 078 b| xamnhe ca 2 mat con 2 thanh kia hdi bi xam md d mQt vai mat.

Nhu" vay hien ti/Ong xam van con xay ra tuy toe d<? con cham hdn co le vitrong nude 16 ft bQt khf.

78

Ngay 29/12/83 nap them 12 BNL nufa va kh6ng cho nirdc vong 1 hoat d<?ng denan che kha nang khong khf lot vao be 16. Ngay 31/12 rut 10 trong so 12 thanh chlde lai 2 thanh so 150 va 152 tai cac 6 7-4 va 7-8. Den ngay 4/1/84 xem 2 thanh 150va 152 (dat ngay 29/12) thay hdi bj xam nhe, xem 3 thanh 078, 180 va 187 (dat ngay21/12 va da thay chung bj xam nhe) thay moi thanh deu co vai mat bj den bong.

Nhu vay hien tu"0ng xam nhien lieu van tiep tgc mac du khong cho nUdc vong1 hoat dong.

d) Ngay 21/12/83 da gCil sang Moskova 2 thanh nhom chen bj xam va 1 chairvJdc lay tu1 be 16. Ngay 12/1/84 nhan di/dc ket qua phan ti'ch ni/dc : pH = 6.2, dp dan= 0.5 micro Sm/cm n6ng do (ddn vi micro g/L) cua C\ = 0, Al= 30, Fe = 50. Cac ketqua nay phu hop vdi cac ket qua phan tfch tai Da lat va cho thay rang chat lUdngnude trong be 16 dat cac yeu cau doi vdi vi#c su" dung loai nhien lieu cua Lien X6.(Hdn nCra, phia Lien X6 se thong bao la d Moskava) da thanh lap hoi dbng 10 chuyenvien de xem xet cac thanh nhom bj xam va ket luan la vdi hien tudng xam nay van cothe tiep tuc du"a \6 vao hoat dpng.

Ngay 12 va 13/1/84 rtta sach cac bQ IQC CO va mQt bo loc trao doi ion cua h#nude vong 1. Ngay 14/1/84 bat dau nap nhien lieu vao vung hoat de tao cau hinh lamviec.

XII. TAI LIEU THAM KHAO

I'M Pham Duy Hien, Ngo Quang Huy, Vu Hai Long, Tran Khanh Mai. Bao caokhdi dpng L6 Phan Orng Hat nhan Da Lat, Phan 1 " Khdi dong vat ly vdi cau hinhvung hoat khong co bay ndtron",

121 Arhanghenski N.V, Bien ban khdi dpng L6 Phan ufng Hat nhan Da Lat(tieng Nga), 1984.

79

VN9700004

BAOCAOKH6I OONG LO PHAN L/NG HAT NHAN DA LAT

PHAN 3

KHCJl DQNG NANG L.U0NG(6/2/1984-15/2/1984)

Pham Duy Hien, Ngo Quang Huy, Vu Hai Long, Tran Khanh Mai

Sau khi hoan thanh khdi dpng v|tt ly 16 phan ufng vdi cau hinh vung hoat 88Bo Nhien lieu (BNL), ngay 6/2/1984 bat dau khdi dpng nang li/png. Cong suat Iddat 10KW (6/2/1984), 100KW (7/2/1984), 200KW va 300KW (8/2/84), 400KW va500KW (9/2/84). TCr 19hO6 ngay 11/2/84 den 19hO6 ngay 14/2/84 cho Id hoat dpnglien tuc 72 gid d cong suat danh djnh 500KW.

Bao cao trinh bay cac ket qua do dac trong qua trinh khdi dpng nang lu*0ngxac dinh cac dac trifng sau day:

- Chuan cong suat 16,- Cac thong so cong ngh# va an toan bufc xa,- Mat do thong lU0ng ndtron,- He so nhiet dp cua dp phan Crng,- Dp nhiem dpc Xenon.

). CHUAN CONG SUAT L6

Cong suat nhiet cua 16 do bang cong suat tai nhi#t cua h# thong vong 1 vavong 2. Tuy nhien cong suat nhi#t chl do dirtfc chinh xac tu" 100KW trd len. Viecchuan cac mure cong suat thap di/a vao phep do thong \U0ng ndtron tuy^t doi.

Viec chi thj cong suat 16 the hi#n tren ban di'eu khien, do h# AKNP tienhanh. He AKNP gom 9 buong ion hoa ghi dong ndtron va he xti ly tfn hieu tCf 9buong ion hoa do de cho biet mtfc cong suat 16.

Cong suat di/dc chia lam 3 dai, moi dai dung 3 bubng ion hoa nhu nhu* dupetrinh bay tren bang 1.1. Trong bang nay N la cong suait danh djnh 500KW. Van decan giai quyet trong qua trinh chuan cong suat 16 la lam sao de cong suat thu"c cua16 trung vdi cac gia trj cong suat chl thj trong cac dai do neu tren. Muon vay, d cacmure cong suat Dl va PD ta dung phep do mat dp thong lirpng ndtron nhiet tai motvai vi tri trong vung hoat ro\ tinh ty l# tifdng doi vdi mat dp thong li/0ng ndtron nhiettheo tinh toan thiet ke tai cac vj trf do. Sau khi so sanh xong ta dat mufc cong suatchf thi phu hpp vdi mufc cong suat that. Qua trinh nay tien hanh den mufc cong

81

suat vai chuc KW. Sau do vdi cac mufc cong suat cao hdn dung phep do cong suatnhiet de chuan dinh cac mure cong suat chi thi.

Bang 1.1. Cac dai do cong suat 16.

Dai

Dai

Dai

nguon

trung gian

nang li/dng

Dai

1E-8

1E-3

1

cong suat

- 1 E-2 %

- 10 %

- 120 %

N

N

N

Ky hi#u

DM,

PD1,

ED1,

buong

DI2,

PD2,

ED2,

ion hda

DI3

PD3

ED3

The loai

KNK-15

KNK-15

KNK-3

1.1- Chuan bang thong li/oYig ndtron nhi#t

Be thuan tien trong viec chuan cong suat ta dung he do bo sung g*6m buongion hoa KNK-56 dat trong kenh ngang so 4 va thiet b| do dong cua buong nay lamay ty ghi KSPV-4. May nay co 11 dai do, moi dai co 100 dp chia vdi mure ddngcu"c dai d moi dai nhu sau:

Dai doDong do CL/C dai, nA

10,2

20,3

30,6

41

52

65

710

820

950

10100

111000

Ta dung he do nay cho so li#u trung gian de dieu chinh mufc cong suat chfthi theo mure cong suat 6U0c xac djnh bang thong Itfdng ndtron nhi#t. Mat 6Qthong lUdng ndtron nhiet dUdc xac djnh bang phi/dng phap kfch hoat la do Aukhong bQC Cd va boc Cd. Phan ufng kich hoat la Au-197 (n,y) Au-198, vdi tiet dienkfch hoat 98,7 ± 0,2 barn. Nguyen to Au-198 cd thdi gian ban ra 2,696 ± 0,002ngay, nang lu"dng bufc xa gamma du"dc do la 412keV vdi h# so phan nhanh95,48%. Sau day la cac Ian chuan cong suat vdi cau hinh 88 BNL.

- Ngay 17/1/1984 - Diia 16 le"n trang thai tdi nan vdi v| trf cac thanh dieukhien la 4KC = 447 mm, AR = 300 mm. Chieu la do Au tai tarn bay ndtron. Mat dothong lUdng ndtron nhiet do di/dc la 2.3E8 n/crr»2.s . Theo tfnh toan thiet ke mat dpthong lifdng ndtron nhiet tai bay ndtron d cong suat 500 kW la 1.8E13 n/cm2.s dodo cong suat tu"dng doi do dUdc la P = 2.3E8/1, 8E13 = 1.17E-3 %N, cong suattuyet doi la P = 5,81 W.

May KSPV-4 6" thang do 11 6Q chia 48 ufng vdi dong do laI = 48/100 x 1000 nA = 480 nA.

Chi thi cong suat tren ban dieu khien nhii sau:

82

PD1 = 1.0E-1 %N DM = 1.6E-1 %NPD2 = 7.9E-2 %N DI2 = 1.6E-1 %NPD3 = 2.5E-1 %N DI3 = 5.0E - 2 %N.

Ta thay rang d cac dai PD va Dl cong suat chl thi cao hdn cong suat thi/cden 2 bac.

- Ngay 19/1/1984 - Rut bubng ion hoa KNK-56 d kenh so 4 ra mot doan saocho dong cua KSPV-4 giam di 10 Ian. Bi/a 16 len tdi han, vj tri cac thanh dieu khienla 4KC = 440mm , AR = 300mm. Mat dp thong li/0ng ndtron nhiet tai bay ndtrondo bang kich hoat la do la 2E9 n/cm2.s.

Cong suat tircfng Cfng la P= 2E9/1.8E13 = 1 , 1 E - 2 % N = 55W.May KSPV-4 6 thang do 11 dp chia 56 Ofng vdi dong do la I = 560 nA.Chi thi cong suat tren ban dieu khien nhif sau:

PD1 = 1,0 %N DJ1 = 2.5E-1 %NPD2 = 1,0 %N DI2 = 1.6E-1 %NPD3 = 2,5 %N DI3 = 3.2E-1 %N.

Ta cung thay rang cong suat chl thj cua cac dai PD va Dl cao hon congsuat thi/c (100 fan va 10 Ian) .

Cung vdi la do chieu tai bay ndtron con chieu la do tai 6 7-1. Mat dp ndtrondo du"dc la 8.95E8 n/cm2.s, tinh toan thiet ke tai vj tri nay Lfng vdi cong suat 500kW cho thong li/dng bang 7E12 n/cm2.s, do do cong suat tu"ong tfng la P = 1,28E-2 %N = 64 W.

Nhu vay hai ket qua do mat dp thong lUOng ndtron nhi#t tai bay va 6 7-1 chocac gia trj cong suat phu h0p nhau va khong phu hpp vdi cong suat chi thi.

- Ngay 21/1/1984 - Rut bubng ion hoa KNK-56 6 kenh so 4 ra co 50cm degiam dong cua KSPV-4 di 10 fan. Bat tdi han vdi 4KC=445 mm, AR = 300 mm.Mat dp thong Itfdng no'tron do 6\J0c tai bay 4,19E10 n/cm2.s, tai mam quay 3.58E9n/cm2.s. Tfnh toan thiet ke tai mam quay cho 3E12 n/cm2.s d 500 kW, do do suy racong suat

P = 2,33E-1 %N = 1164Wtaibay,P = 1.19E-1 %N = 597 W tai mam quay.

Nhu" vay hai ket qua khong trung nhau: d bay 1,2kW , d mam quay 0,6kW.Tuy nhien nhin vao vung hoat Ian dau tien thay anh sang bufc xa Xerenkov.

Do bubng ion hoa KNK-56 d kenh so 4 rut ra de dong cua KSPV-4 giam di10 Ian, trong luc do cong suat tang gap 10 fan so vdi ngay 19/1/1984 nen du* doanKSPV-4 se chl gia tri gan nhu" ngay 19/1/1984. Thi/c te la d thang do 11 dp chia 80dong do la 800nA.

De cong suat chl thi phu hop vdi cong suat thu"c, da phai tien hanh dfeuchlnh v| tri 3 bubng ion hoa dai PD cua h# AKNP sao cho cong suat chl thj cungvao cd 0,1 %N. Sau khi dfeu chlnh, chl thj cong suat tren ban di'eu khien la

PD1 = 1,3 E-1 %N ; ED1 = 0,5 %NPD2 = 2 E-1 %N ; ED2 = 0,1 %NPD3 = 1,3 E-1 %N ; ED3 = 0,4 %N

83

Nhu" vay chf thi theo dai PD phu hdp hdn mufc cong suat cong suat 16.

- Ngay 23/1/1984 - Kiem tra he thong dieu khien tu" d<?ng AR, xem cong suatId co tf le vdi cong suat thiet lap theo AR hay khong. Cong suat 16 dUdc di'eu khientu" dgng theo cong suat thiet lap theo AR nhd tfn hi#u lay d bubng ion hoa dai PD.

Bi/a 16 len tdi han vdi vi trf cac thanh dieu khien 4KC - 454 mm va AR =300 mm.

Cong suat thiet lap theo AR = 0,5 %N. Vdi gia tri KSPV-4 tai thang do 4 dochia 58 dong do la 0,58 nA. Tang cong suat theo AR len 1% thi cong suat tang vaso do KSPV-4 tai thang 5 dp chia 62 cho dong do bang 1,24 nA. Ta thay rang tf lecac dong do la 1,24/0,58 = 2,14. Tang cong suat theo AR len 5% thi cong suat tieptgc tang va so do KSPV-4 tai thang 7 do chia 62 cho dong do 6,2nA.Ta thay rang tfle dong do la 6,2/1,24 = 5 . Nhu" vay dong KSPV-4 tang tl l# vdi cong suat thiet laptheo AR. Di'eu do chufng to he thong dieu khien tu1 d<?ng lam vi#c tot.

- Ngay 24/1/1984 - Tiep tuc kiem tra he dieu khien tu" dQng AR nhd tin hieulay tu1 buong ion hoa dai PD.

Cong suat thiet lap theo AR = 1%.Dong KSPV-4 : Thang 6 6Q chia 35 = 1,75 nA.Cong suat thiet lap theo AR = 44,4%Dong KSPV-4 : Thang 10 6Q chia 96 = 96 nA.Tf le ddng = 54,8.Tf le nay khong trung vdi tf le cua cong suat thiet lap theo AR. Co the do hai

gia tri cong suat dat qua xa nhau.

1.2- Chuan bang cong suat nhi#t

- Ngay 6/2/1984 - Bat dau khd'i dpng nang Itfdng. Rut bubng ion hoa KNK-56ra khoi kenh so 4. Du"a 16 len tdi han vdi v| trf cac thanh dieu khien la 4 KC = 458mm, AR = 300 mm.

Ou"a cong suat len cd 10 kW bang each chon cong suat chi thi cua dai PDbang 2%. Khi do chi thj cong suat la

PD1 = 2 %N ED1 = 3 %NPD2 = 2 %N ED2 = 3,6 %NPD3 = 2 %N ED3 = 2,8 %N

- Ngay 7/2/1984 - Trong ngay 21/1/1984 cong suat chf thi theo dai PD phuhdp vdi cong suat thi/c cua 16, con cong suat chf thj cua dai ED khong phu hdp.Ngay 23/1/1984 kiem tra he thong di'eu khien tu* dong AR nhd tfn hieu tu* bubng ionhoa dai PD thay rang gia \:\ cong suat thiet lap theo AR tang tf le vdi cong suatthu"c cua 16 do bang may KSPV-4, song mufc cong suat thiet lap nay chua phu hdpvdi cong suat chi thi cua dai PD. Dieu nay xay ra do cong suat thiet lap theo ARtheo dai PD Idn hdn cong suat thu"c cd 10 Ian. Vi vay can chuan cong suat thiet lap

84

theo AR trong dai ED cho phu hqfp vdi c6ng suat chl thj, tufc la phu hop vdi congsuat thu"c cua 16.

Daa 16 len tdi nan dat mufc c6ng suat 0,05% theo dai PD. Bat cong suatthiet lap theo AR = 1% theo dai PD th) cong suat 16 tang va di/o*c gift 6 mufc 0,1 %theo dai PD. Chuyen cong suat thiet lap theo AR len 5% thl c6ng suat 16 tang vadu"0c gift of mftc 0,5% theo dai PD. Nhu1 vay hi#n tai c6ng suat 16 la 0,5%. Chuyentin hieu dfeu khien cong suat tft dai PD sang dai ED, tufc la dung t(n hieu tft buongion hoa dai ED de so sanh vdi c6ng suat thiet lap theo AR. Dfeu chlnh h$ sokhuyech dai cua cac so1 do khuyech dai cua cac buong ion hoa dai ED sao chocong suat chf thi dai ED bang 0,5%, khi do nhan duXJc cong suat trung binh cua daiED la 0,32%. Khi chuyen cong suat thiet lap ve dai ED cung dat tai gia trj congsuat theo AR bang 0,5%.

Sau khi chuan cac cong suat chl thj ED va PD trung vdi cong suat thiet laptheo AR, bat dau tang cong suat thiet lap theo AR va nhan du"o"c cac mtifc congsuat 16 nhu sau (Bang 1.2)

Bang 1.2. Cac mu"c cong suat 16

C O N G S U A TAR (%)

15

1020

TB (%)0,634,59,7

19,9

ED1 (%)

1,14,99,8

19,4

ED2 (%)

1,16,0

11,823,7

ED3 (%)1,04,69,1

18,3

NHIET(KW)5

2550

100

Nhir vay cong suat 16 dat 100 KW theo chi thj cua dai ED va cong suat nay6U0c gift nho" he thong dieu khien tu" dong AR. Vi tri cac thanh dieu khien khi 16 6'mufc cong suat 100KW la KC1 = 457 mm, KC2 = 426 mm , KC3 = 457 mm , KC4= 468 mm , AR = 305 mm.

Tinh cong suat nhiet cua 16 khi 16 lam vi#c 6 mufc cong suat 100KW. Hinh 1trinh bay cac diem do nhiet d<? trong be 16 va he thong nude tai nhiet. Cong suatnhiet 6U0c ti'nh nhu1 sau. Doi vdi vong 1 dung nhiet do ni/dc loi vao Tv va nhiet donu"dc loi ra Tr cua binh trao doi nhiet de ti'nh cong suat nhiet cua vong 1 theo congthftc

P1 = 1,16xG1 x (Tv-Tr) (KW) (1-1)trong do G1 la li/u lUOng nJdc vong 1, m3/h. Doi vdi vong 2 ta cung co cong thufctinh cong suat nhiet la

P2 = 1,16 x G2x (Ur-Uv) (KW) (1-2)trong do G2 la lUu \U0ng nt/dc vong 2, m3/h con Uv va Ur la nhiet do ni/6'c loi vaova loi ra cua binh trao doi nhiet 6 vong 2. Sau day la cac so lieu do di/o'c va tinhtoan theo cac cong thtfc (1-1) va (1-2). Trong thi/c te khong do Tv va Tr ma do tri/ctiep DT = Tv - Tr (Bang 1.3).

85

m3/h GTT= 90 m3/h

II

T.

>i.n.a 1 . ;>>• do tai nhiet va cac diem do nhiet do

86

Bang 1.3. Cong suat nhi^t cua 16.

Thdi diem

14h20

14h45

15h30

16h15

17h00

DT°C

0,92

1,4

1,50

1,56

1,58

G1m3/h

49

49,5

49,5

49,5

49,5

P1kW

52,3

80,4

86,1

89,6

90,7,

UvOC

14,3

14,0

13,5

13,2

13,0

Ur°C

14,6

14,5

14,2

14,0

13,7

G2m3/h

91

91

91

91

91

P2kW

31,7

52,8

73,9

84,4

73,9

Theo bang 1.3, cong suat nhiet cua ni/dc vong 1 va vong 2 tang dan theothdi gian va dat P1 = 90,7 kW , P2 = 84.4KW. Do sai so do nhi#t dp vao cd 0,5°Cnen cac gia tri cdng suait tfnh tren co sai so Idn c6 30% doi vdi P1 va 50% doi vdiP2. Mat khac do che dp nhi#t trong 16 chUa on djnh nen cd the cong suat nhietchUa dat gia tri can bang.

- Ngay 8/2/1984 - Bite Id len den cdng suat 300 KW. DUa Id len tdi nan vadJa len cong suat 200KW ( P AR = 40 % , ED1 = 37,8 % , ED2 = 46 % ,ED3 = 36 %, P TB = 39,5 % ). Vj tri cac thanh dfeu khien la 4KC = 445 mm ,AR = 325 mm.

Sau 30 phut cac thong so nhiet la

G1 = 49,3 m3/h , Tv - TrG2 = 94,2 m3/h , Ur - Uv

1,88°C , P10,8°C , P2

107,5 KW87,4 KW

Bi/a cdng suat len 300 KW (P AR = 60 % , ED1 = 58 % , ED2 = 70,5 %ED3 = 55,1 %) V| tri cac thanh dieu khienla 4KC = 448 mm , AR = 175 mm.

Cdng suat nhiet cua cac vdng tuan hoan tang theo thdi gian nhu" tren bang1.4.

TO bang 1.4 thay rang cdng suat nhiet vdng 1 dat gia trj 300KW con cdngsuat nhiet vdng 2 thap hdn dat 253KW.

87

Bang 1.4. Cong suat nhiet tang theo thdi gian.

Thdi diem

11h2011h5012h2012h5013h2013h5014h20

DT°C2,7

3,634,184,324,464,704,82

G1m3/h48,349,549,449,549,449,549,2

14h46 L6 bi dap, sau do len lai15h1516h0016h3017h0017h30

3,834,564,784,985,08

49,450,650,650,650,6

P1kW151210240248256270275

cong suat G1,82,02,22,12,3

Ur-Uv°C1,01,41,61,51,71,81,8

I00KW luc1,82,02,22,12,3

G2m3/h94,594,894,894,895,095,095,0

15h95,095,095,095,095,0

P2kW110154176165187198

198

198220242231

253

- Ngay 9/2/1984 - Baa 16 len cong suat 500KW. Kiem tra he thong dap Idtheo tin hieu AZ.

Bu"a 16 len tdi nan va dat cong suat 20KW (P AR = 4 %). D£p 16 bang eachbam nut AZ, cac thanh AZ , KC va AR rdi xuong tan cung.

Du"a Id len cong suat 300kW (P AR = 60%). Dat mite su* co theo cong suatla 70% . Dap Id theo tin hieu giam lull lU0ng niidc vdng 1. Tat ca cac thanh dieukhien deu rdi xuong tan cung.

Bi/a 16 len cong suat 400KW ( P AR = 80 % ). Bi/a tiep len 500kW ( P AR =99,9%). Bat mu"c su" co theo cong suat la 110% . Dap 16 theo tin hieu vi/dt mufccong suat. Tat ca cac thanh dieu khien deu rdi xuong tan cung.

- Ngay 10/2/1984 - BUa cong suat len 500KW va xac dinh cong suat nhietcua vong 1 va vong 2. Cong suat 500 KW dat luc 10h10 (Bang 1.5).

- Ngay 11/2/1984 - Luc 19hO6 cho Id bat dau hoat d<?ng d cong suat danhdinh 500KW va keo dai trong 72 gid lien cho den 19hO6 ngay 14/2/1984. Congsuat nhiet dU0c tfnh theo cac gia tri do di/dc cua cac thong so nhi#t ky thuat vdng1 va vdng 2.

88

Bang 1.5. Cong suat nhi#t tang theo thdi gian.

Thdi diem

10h20

10h5011h2011h50

12h30

13h00

13h30

14hOO

14h3015h0015h30

16h0016h30

17h00

DT°C

5.167,007,457.788,008,208,208,259,229,709.569,339,209,15

G1m3/h

5049,849,449,449.449,449,549,549,549,550,049,550,049,5

P1kW

299404427446458470471474529557554536534525

Ur-Uv°C

2,23,03,13,23,23,33,43,34,03,93,93,63,73,7

G2m3/h

93,594,094,094,094,094,093,593,593,593,593,593,093,593,5

P2kW

239327338349349360369358434423423390401401

Chu y : luc 14hOO bat cac quat 1-1 va 1-2 d thap lam nguoi.

Cong suat nhiet vong 1 tang dan va dat 500KW luc 20h30 ngay 11/2/1984.Sau do tang den 550KW luc 2h00 ngay 12/2/1984, luc bay gid Tv-Tr = 9,5°C . Saudd'hieU nhiet do nay tang len den 10°C (13h36), cong suat dat 580 KW. Luc8h40 ngay 12/2/1984 (gid thtf 14 ke tu" thdi diem ban dau) giam cong suat xudngcon 90%, tile con 520 KW.

- Ngay 13/2/1984- Luc 6h00 sang tien hanh do nhiet 6q> vdng 1 Tv va Trrieng biet de ti'nh hieu Tv-Tr va so sanh vdi ket qua do DT tru"c tiep. Thay rang Tv-Tr = 28,76 - 22,57 = 6,19 °C thap hdn nhieu so vdi gia trj do tri/c tiep DT = 9 oc.Nhi/ vay cong suat thi/c te ufng vdi DT= 6,1.-9 °C la 360 KW, ch? dat 72%cong suatdanh dinh. Do do can phai tang cong suat 16 len den cong suat danh dinh. Muonvay, luc 13h21 ngay 13/2/1984 di/a lai cong suat thiet lap P AR = 99,9 %. Luc14h25 (gid thu" 43 ke tu1 thdi diem ban dau) dieu khien cong suat theo ED1 con cacmifc ch! thi cong suat cua ED2 va ED3 dat lai la 70%. Luc 14h30 dat lai cong suatthiet lap P AR = 70% va di'eu khien cong suat theo ED2 (luc do ED2 = 70%),chuyen ED1 chi thi 70%. Nhu1 vay cong suat mdi 70% ufng vdi cong suat cu 100%.Sau do'tang d'an mufc cong suat thiet lap P AR len 80%, 90% roi 99,9% luc 17hO6(gid thuf 47). Cong suat 16 dat mure 100% danh dinh , so vdi trade khi di'eu chinhtang 100/70 = 1,43 Ian. Sau khi dieu chlnh, tiep tuc theo doi cong suat cua he vong

89

1 bang each do nhiet 66 rieng biet Tv va Jr. Ket qua cho thay hieu so Tv-Tr daodpng trong khoang 8,7 °C den 9,3 °C, nghTa la cdng suat dao dong trong khoang505 KW den 540 KW. Cong suat tfnh theo vong 2 dao dpng trong khoang 470 KWden 522 kW do hieu so nhiet dp Ur-Uv dao dong trong khoang 4,5 °C den 5 °C.

Nhu vay, sau dpt chay 16 72 gio lien cdng suat 16 dupe chuan lai de congsuat chi thi tren ban di'eu khien, cong suat thiet lap theo AR va cong suat nhiettrung nhau.

II. CAC THONG SO CONG NGHE VA AN TOAN BLfC XA

2.1. Nhiet 6Q ni/cfc trong be 16 va cac vong 1, 2

Tren hinh 1 trinh bay sd do tai nhi£t va cac diem do nhiet dp. Nhiet dp Id'ivao va loi ra binh trao doi nhiet cua cac vong nadc tuan hoan du"dc do bang cacnhiet ke dien trd TCM-5071 con nhiet dp tai cac diem trong be 16 tCf T1 den T5 donhd cac nhiet ke dien trd TCP-5076. Cac gia tri nhi#t dp du*dc xac djnh trong dpt72 gid lam viec lien tuc d cong suat 500 KW. Nhi£t dp tai cac diem do thay doitheo nhiet dp moi trudng, thtfdng cao vao ban ngay. Sau day la cac dai nhiet dp dodupe

Nhiet dp loi vao vung hoat T1 = 27 - 30 °CNhiet dp loi ra vung hoat T2 = 39 - 41 °CNhiet dp d phan tren be 16 T5 = 34 - 36 °CNhiet dp d phan giQa be Id T3 = T4 = 26 - 28 °CNhiet dp nude vdng 1 loi vao binh trao doi nhi£t Tv = 34 - 37 °CNhiet dp nude vong 1 loi ra binh trao doi nhi^tTr = 2 5 - 2 8 °CHieu so nhiet dp vdng 1 - 8,7 - 9,3 °CNhiet dp nude vdng 2 loi vao binh trao doi nhi#t Uv = 16,5 - 20 °CNhiet dp ni/dc vdng 2 Id'i ra binh trao doi nhi#t Ur = 21,5- 24,5 °CHieu so nhiet dp vdng 2 = 4,5 - 5 °C

2.2. Cac thong so cua nUcfc vong 1

Dp dan dien rieng trade phin Ipc 0,33 - 0,71 micro Siemen/cm.Dp d in dien riengsau phin Ipc 0,24 - 0,45 micro Siemen/cm.Dp pH cua nude vdng 1 tri/dc va sau phin Ipc 5,5 - 6,2.Ltfu IcTdng nude trong h# thong Ipc made vdng 1 la 2 - 2,2 m3/h.

2.3. Tmh hinh phdng xa

Tinh hinh phdng xa gamma va ndtron trong gian nha Id dadc do trong thdigian Id lam viec 72 gid. Sau day la cac mufc lieu gamma, mat dp thong ladng ndtronnhiet va ndtron tren nhi#t tgi mot so vj tn (Bang 2.1).

90

Bang 2.1. Cac mufc lieu gamma, mat dQ thong \\J0ng ndtron nhi#t va ndtrontren nhiet tai mot so v| tri

V| tri do

Gan thanh betong bao veTren be chufa nhien li#uTren nap loTai lo 6 tarn nap loSat mat ni/dc d tarn nap loTrong phong 148Sat cot trao doi ionTrong phong dieu khien

Suat lieu Gamma(micro R/s)

0,08-0,240,1 - 0,320,03 - 0,3

825

0,1 - 1,2400

0,01 - 0,04

Mat do thong lu"0ng, n/cm2.sNhiet Tren nhi$t

0,05 - 1,53 0,05- 3,50

0,05 - 0,2 0,05- 1,3

26 6,9

Hoat dQ phong xa beta tong cua can nude lo do 2 gid sau khi lay mau nudeId la (3,5 - 6,5) E-7 Ci/I.

Thanh phsm dong vj cua ni/dc lo tai thdi diem lay mau the hi#n tren bang2.2.

Bang 2.2. Thanh phan dbng vj cua ntfdc lo tai thdi diem lay mau.

Dong vj

Al -Mg

ClMn

NaA r -

Xe-

28- 2 7

-38- 56- 2 441

135

Hoat dQ ,Tri/dc phin IQC

0,1 - 0,64 - 6

0,03 - 0,140,12 - 0,190,42 - 0,86

1,6 - 4,4-

micro Ci/ISau

1,8(5,6 -

phin

----- 3

13)

IQC

E-3

C i / h .Hoat do phong xa cua khf phong xa Ar-41 tai ong thai khi la (2,5 - 4,2) E-2

Tong hoat do beta cua khf thai tif ong xa la 1,1E-8 C i / h hay 3,5 E-15 Ci/I.

Trong thdi ky khdi dong vat ly va khdi dQng nang lUOng da kiem tra lieu canhan cho 44 ngUdi trong thdi gian 3,5 thang. Lieu hap thu phan bo theo nhom

di nhu" tren bang 2.3.

91

Bang 2.3. Lieu hap thu phan bo theo nhom ngtfdi.

Lieu hap thu, mR

So ngUdi

0 - 20

6

20 - 40

12

40 - 60

16

60 - 80

7

80-100

3

III. MAT OQ THONG LU0NG NOTRON

Trong bao cao khdi dpng vat ly 16 phan 2 da trinh bay phan bo ttfdng doi cuamat dp thong li/0ng ndtron nhiet theo chieu cao va theo ban kfnh vung hoat. Trongbang 3-1 trinh bay cac ket qua do mat dp thpng ItfOng ndtron nhiet va ndtron nhanhtai mot so vi tri do vdi cac mufc ccng suat khac nhau cung vdi he so Cadmi Red valoai detectd dung.

Mat dp thong Itfdng ndtron nhiet tai bay ndtron d cong suat 500 KW la 2,1E13 n/cm2.s.

IV. HE SO NHIET DQ CUA DQ PHAN

He so nhiet dp cua dp phan ting du*dc xac djnh ngay 21/1/1984 trong thdigian khdi dpng vat ly. Cac ket qua di/dc viet trcng bap cap nay de thuan tien choviec trinh bay hi#u chlnh cac phep do dp nhiem dpc xenpn.

Tri/dc khi tien hanh thi nghiem, nang nhiet dp ni/dc trpng be Id \d 20°C len30°C bang each chc chay bdm vpng 1 khoang 20 gid va khong cho chay bdmvong 2. Nang lu*dng sinh ra trong phan thtiy dpng cua bdm, cd 10KW, dUOc bienthanh nhiet nang va thep he thong ni/dc vong 1 lam nong nu"dc 16.

De do he so nhiet dp, du"a 16 len trang thai tdi nan d mufc cong suat cd 7,9E-3 %N = 40 W. Sau dp chc hpat dpng bdm vpng 2 de ha nhiet dp trpng be 16. Xacdinh thay doi nhiet dp T1 cua Ip'i vap vung hpat theo v| tri thanh AR. Bang 4.1 daytrinh bay cac ket qua do va dp phan ufng dua tren dac trtfng tfch phan cua thanhAR.

Hinh 2 trinh bay si/ phu thupc dp phan uYig vap nhiet dp T1 . Su" phu thupcnay cc dang tuyen tinh

p= aT + b (4.1)trcng dp a la he SP nhiet dp cua dp phang ting.

TO" hinh 2 tinh dUdc he so nhiet dp cua dp phan ufng bang

a = -8E-3 $/°C

92

Bang 3-1: Mat dp thong Itfdng ndtron tai cac vj trf trong 16

Vj trf do

Bay ndtron

Kenh so 3

Kenh so 1

Kenh so 2

Kenh so 4

Kenh 7-1

Kenh 13-2Mam quay

Cot nhiet

Cot nhi#t,hop Cdday 1mm

Thong li/0ng, n/cm2.sNhiet

8,18 E72,3 E82 E9

4,19 E105 E111,64 E111,22 E71,6 E82,08 E10

2,77 E10

1,26 E10

7,14 E10

8,95 E8

2,51 E111,21 E93,58 E99,76 E101,98 E 9

2,63 E8

Nhanh

1,48 E8

6,7 E105 E10

1,8 E8

(5,7 E8 ^2,9 E8

5,25 E7(1,5 E8 ke

1,88 E9(1,135 E1C

1,37 E84,9 E10

2,98 E91,15 E6

Red

3,62,6

2,5

2,72,522,466,75,5

3,639

<et qua c4,27

436,6

Cong suat

wn

210

1001001000

200005000

10100

153501535015350

o tren may F153501535015350

Detectd

AuAu

Au

SAu

Au/SAu/S

AuAu

AuCuS

>SO-2-4)Au/SCu

Au/SI qua do tren may PSO-2-4)

822,2

) ket qua152,0

2,52,93,4

4,6

29,6( 3,5 E6 ket qua do 1

2,37 E8250

1535015350

do tren may15350

100100

200001001000

2000015350

CuAu/S

PSO-2-4)CuAuS

Au/SAuAu

Au/SAu/S

ren may PSO-2-4 )15350

500000Cu

Mn/ln

Ngay do

26/12/8317/1/8419/1/8425/1/8421/1/849/2/8415/2/8417/1/8419/1/846/2/846/2/846/2/84

6/2/846/2/846/2/84

6/2/846/2/84

6/2/8419/1/8425/1/849/2/8419/1/8421/1/849/2/846/2/84

6/2/8411/2/84

O Gia tri cong suat cho 6 day chtfa dUOc chuan lai , xem ph'an 1.2

93

Hinh 2. Sir phu thu§c do phan u*ng vao nhi^t do

loi vao vung hoat.

94

Bang 4.1. Cac thong so khi do h$ so nhi#t dQ.

Thdi diem9h329h369h419h469h519h5610h0110h0610h1110h16

T1,°C30,0

27,826.225,324.523,823,222,822,6

AR, mm203

cho chay bdm vdng 2225238246250254257260-263

Dp phan tfng, $0,102

>

0,1200,1340,1420,1470,1510,1550,1580,161

V- B 0 NHIEM DQC XENON

Trong qua trinh lam vi#c cua Id phan ufng, Xe-135 du"dc sinh ra tti san phamphan hach, no hap thu manh ndtron nhi#t va do do di/a vao dQ phan ting am. BegiuTIo lam viec d mtic cdng suat co djnh can phai rut cac thanh dieu khien len debu trCr do phan ting am nay.

TO1 he phUdng trinh vi phan mo ta qua trinh dong hoc cua hi#u ting nhiemdoc Xenon suy ra bieu thtic sau day cua nong do" hat nhan Xe theo thdi gian vathong Itfo'ng ndtron trung binh <j> .

X(t) = exp [-X x*(<j>) t] {JQ1 [ px Z\J <j> + P| Zuf <j> (1- exp(-X| t)) +

+ IOAJ exp(-X| t ) ] exp ( X x*(<j>)t )dt + Xo} (5.1)

trong do Xo va l 0 la nong do hat nhan Xe va lot 6 thdi diem ban dau; hang so phanra tudng diio'ng

X V T / = X "*" ^ X V • W - * - /

Bo nhiem doc Xenon phu thuoc vao thdi gian lam viec cua Id va mat dothong li/dng ndtron trung binh nhu" sau:

a x X (t)

PxO) = - f- = - Px N(t) (5.3)

Vdi do nhiem doc Xe can bangy f

P°x ( Px + Pi ) (5.4)

95

Cac so lieu hat nhan su" dung trong cac bieu thufc tren nhu" sau:p, = 0,056Px = 0,003X, =2,8E-5/sec = 0,1008/h,X x = 2,09E-5/sec = 0,0752/h,cr x = 2.75E6 barn (d nhiet do cua vung hoat)

I uf / I u = 0,85 doi vdi U-235 36%.f = 0,853 ( he so su" dung ndtron nhiet)<j> = x.1E12n/cm2.s

Trong trirdng hop n'ong d<? Xe va lot d thdi diem ban dau bang khong va vdigia thiet thong Itfdng (j> trong (5.1) khdng phu thuoc t, tfch phan (5.1) cho bieu thCrcddn gian sau day cua ham N (t) trong (5.3)

N (t) = {1 - exp [-X x*(<j>0) t] } - p k((j>0) {exp(-X, t) - exp[-Xx*(<|>0) t ] } (5.5a)trong do

P = P1/(P1 + Px)k = k (cj)0) = Xx *(*) / [Xx *(•) - X, ]

Trong trudng hdp nong do Xe va lot d thdi diem ban dau khac khong, hamN(t) dt/dccho bdi :

N (t) = [ 1 - e - ^ - x ^ ^ ) J] +

- • o+ p k(^)[ (1 - e " x I to) - 1] [e ~x I * - e " V C ^ ) { ] +

- P-k( <fr0 ) (e- Vo - e'x x*(<b0) tQ)j e^V^-,)* (5.5b)

vdi t tinh tu" thdi diem t0 khi thay doi thong lUdng ndtron id $0 len (j>-j va Xo va l 0

la nong do Xe va lot d thdi diem to :

Pi Suf <t>0

l0 = - (1 - e' Ho ) (5.6a)X,

(Pi +PX) V ^o

0

X0 = (1 - e * V ( * o ) t

96

V - e - V(4>0) *0) (5.6b)

Thf nghiem xac dinh di/dng cong thay doi do nhiem doc Xenon va ti7 do tinhdo nhiem doc Xenon can bang 6\J0c tien hanh trong dot 16 hoat dpng 72 gicf 6' congsuat danh dinh. Viec xac dinh do nhiem doc Xenon dtfOc tien hanh bang phtfdngphap bu trCr.

Trong thdi gian 16 hoat dpng, ngoai do phan ufng am do do nhiem docXenon cac hieu Crng khac nhu" nhiem doc samari, hieu ting tang nhiet do cung du"avao do phan ufng am. Do do si/ thay doi tong cpng cua do phan ting la:

P = Px + Ps + PT (5-7)

tCr do xac djnh difoc do phan C/ng do nhiem doc xenon.Bo phan ufng p j do hieu ufng nhiet dp bang

PT = a ( T - T o ) (5.8)trong do a = -8E-3 $ / C ,

To va T la nhiet dp ban dau va nhi#t dp dang do.

Dp phan urng p s dupe tinh theo bieu thufc sau :

Ps(t) =

i f

n o - . f pPs ~ T r D

cac hang so su1 dung trong bieu thufc tren nhu" sau:

Pp =0,011 , a s =5E4barn , X p = 4,1E-6/sec.Suy ra dp nhi lm doc samari can bang:

p s° = -8,1 E-3 = - 1 $ .

Bang 5.1 trinh bay cac ket qua dp do nhiem doc xenon theo thdi gian trongdo"t lam viec 72 gid tLf 11/1/1984 den 14/2/1984. Cot dau cho thdi gian 16 6 congsuat, cot 2 cong suat 16. 6 day co su1 thay doi lai gia tri cong suat 16 so vdi gia trighi trong nhat ky van hanh 16. Trong nhat ky, vao luc 19hO6 16 dat cong suat100%, den gid thi? 14 chuyen xuong 90%, den gid thur 43 chuyen len lai 100% vavao gid thir 47 nang len den 143%. Cong suat 143% cuoi cung du"0c SLV dung tu"day v'e sau, coi nhu" mire 100%, do do cac mOfc cong suat trudc day, thu'c te thap

97

Ba.G9.-5JL- Ket qua xac d|nh do nhiem doc Xenon

CO00

t(h)151014152025303540444547505560657072

P(%)707070636363636363637070100100100----

KC1(mm)420---

410400---

390

-

-385--

380-

KC2(mm)420---

410405---

395

-

-385--

375-

KC3(mm)420---

410400---

390

-

-

385--

380-

KC4(mm)420---

410405---

395

• -

-

385--

375-

AR(mm)450335200

314400300250195400

335

270400310220357349

P($)

0.000-0.130-0.280

-0.437-0.575-0.695-0.750-0.805-0.904

-0.971

-1.039-1.154-1.259-1.354-1.397-1.402

T

25231.527.2

25.826.426.024.623.025.3

27.3

28.627.827.629.328.527.5

PTL$]

0.000-0.050-0.016

-0.005-0.010-0.006+0.005+0.0170.000

-0.017

-0.027-0.021-0.019-0.033-0.026-0.018

Ps($)

0.000-2 E-4-4 E-4

-7 E-4-1 5E-4-0.002-0.003-0.004-0.005

-0.006

-0.008-0.009-0.011-0.012-0.014-0.015

Px

0.000-0.125-0.264

-0.431-0.564-0.687-0.752-0.818-0.899

-0.948

-1.004-1.124-1.229-1.309-1.357-1.369

hdn 100%. Cu the la trong khoang thdi gian tu* dau den gid thuf 14, cong suat 16 la70%, tCr gid thu" 14 den gid thuf 44, cong suat la 63% , tu1 gid thuf 44 den gid thuf 47cong suat la 70% va gio* thu" 47 tro" di cong suat la 100%. Tren cac cot tiep theo ghivi tri cac thanh KC va AR . De tfnh dp phan ufng tong cpng p ta sCf dung cac dudngcong dac trifng tich phan cua cac thanh KC va AR trong bao cao phan 2 va he sogiao thoa giufa cac thanh KC la 0,88. Qui tfdc p = 0 tai gid thuf nhat de tinh p 6'thdi gian tiep theo. Dp phan u"ng pj di/0c tfnh theo bieu thufc (5.8) vdi nhiet dpI6i vao vung hoat T1. Dp phan Lfng am do nhiem doc Samari di/dc tinh theo bieutntfc (5.9). Cot cuoi cung cua bang 5.1 la dp phan urng do dp nhiem doc xenon6U0c tinh theo bieu tntfc (5.7)

Ket qua xac dinh dp nhiem doc xenon dup"c trinh bay tren hinh 3. Ta thayrang trong 47 gid dau, dp nhiem doc xenon tang dan theo thdi gian. Tu1 gid thu" 47,cong sua't 16 tang tu1 70% len 100%, dp nhiem dpc xenon tiep tuc tang ma chua datden trang thai can bang.

Ta hay tinh dud'ng cong nhiem dpc xenon theo bieu thufc (5.3) va (5.5) va sosanh vdi du"dng cong thi^c nghi^m. TCr do, co the xac dinh di/o'c mat dp thong lifp'ngncftron trung binh trong vung hoat ufng vdi dp nhiem dpc xenon do diTdc.

Ta chu y rang cong tntfc (5.5a) dung trong trUdng hp"p nong dp lot va Xenond thdi diem ban dau bang 0. Do do cong thufc nay dung vdi phan di/dng cong tu"dau den thdi diem 47 gid. Trong khoang thdi gian nay mat dp thong Itfdng ndtron la(j)0 Lfng vdi cong suat 63%. Tu" thdi diem t0 = 47 gid trcf di cong suat tang len100%, do do mat dp thong lu"0ng no'tron la §-\ = 1,59.(j)o . Khi do

crx (|)0 = 0,0099.x/h a x <j>i = 0,0157x/h

^x '(^o ) = (0,0752 + 0,0099.x)/h ^ x *{fa) = (0,0752 +0,0157.x)/h

Tien hanh tinh cac bieu thufc (5.5a) trong khoang thdi gian tu" 0 den 47 gid(5.5b) tu1 47 gid den 72 gid vdi mpt so gia tri cua x , xac djnh dUOc gia trj x = 1,9cho ket qua tinh toan phu hop vdi cac diem thi/c nghiem, xem dudng cong ve trenhinh 3. Mat dp thong lUOng ndtron trung binh trong vung hoat ufng vdi cong suatdanh dinh la:

<|>1 =1,59x. 1E12 n/cm2.s = 3E12 n/cm2.s .

Vdi gia tri nay, suy ra dp nhilm dpc xenon can bang la:

Px °(<t>i) = - 0,01214 = -1,5$

99

ti;

P CDS3 3

O

n-'i

oonro

Q,

o

on

100

VI. KET LUAN

Trong dot khdi dpng nang \\J0ng da dtfa cdng suat 16 len dan dan va datmtfc cong suat danh d|nh 500 KW. Cong suat 16 dtfOc chuan theo cong suat nhi$tcua he thong nude vong 1 va 2. 6 mufc cong suat danh djnh, mat dp thong lUOngndtron nhiet dat gia tri cao nhat la 2,1 E13 n/cm2.s tai bay ndtron.

L6 phan ufng da hoat dpng lien tuc 6" mufc cong suat danh dinh trong 72 gid.Trong dot chay 16 nay da khao sat cac thong so cong ngh? 16 nhi/ nhi^t dp tai cacvi tri trong be lo, nhiet dp nude vong 1 va vong 2, cac thong so nu"dc be 16. Dbngthdi cOng khao sat tinh hinh phong xa tai cac ndi lam vi#c, dp phong xa cua nub'c lova khf thai. Noi chung, cac thong so nam trong gidi han cho phep lo hoat dpng antoan.

Ket qua khdi dpng vat ly va khdi dpng nang li/dng cho ket luan rang Lo phanu'ng hat nhan Da Lat co the dtfOc khai thac an toan d cong suat danh djnh 500KW.

101

CAC CONG TRiNH KHOA HQCNGHIEN CLTU CAC DAC TRLfNG VAT LY LO

CUA LO PHAN LfNG HAT NHAN DA LAT

103 • | — ~ -

I Irff

Proceedings of the Seminar on the Use of Research Reactorsin Fundamental and Applied Sciences

Tripoli, Libya, 16 -20 September 1984, pp. 143-150.

PRELIMINARY EXPERIMENTS PERFORMED ON THENEWLY RECONSTRUCTED 500 KW NUCLEAR

RESEARCH REACTOR OF BALAT

Pham Duy Hien and Ngo Quang HuyNuclear Research Institute,

DALAT, VIETNAM

I. INTRODUCTION

In 1975, all the fuel elements of the former nuclear reactorTRIGA MARK II in DALAT had been dismounted and taken away, whicheffectively caused this atomic research center to stop its activities forsome time. Fortunately, in 1978 a plan for reconstruction of the reactorwith a higher operating power of 500 Kw, twice that of the previous one,was officially approved through technological assistance from the SovietUnion. According to this project, the undismountable components of theformer reactor such as the reactor tank and concrete shielding, the gra-phite reflector, the beam tubes, the thermal column etc. are to be retained,while new design work must be carried out for such new components andsystems as the reactor core using Soviet made fuel elements type VVR-M2,the control system, the cooling system and the whole radio-protection andsafety systems of the nuclear center. The former bulk shielding tank hasbeen redesigned into a storage tank for burnt-up fuel elements. The re-construction work was started in March 1982, and after two years, the newreactor reached its planned nominal power of 500 Kw for the first timein February 1984.

II. SOME FUNDAMENTAL CHARACTERISTICSOF THE REACTOR

A view of the nuclear reactor and a top view of the active coreare presented in Fig. 1,2. At the center of the core is situated the neutrontrap which has a diameter of 65 mm and the wall of which is made of Be.The region between the active core and the old graphite reflector is alsolined with beryllium.

105

o : J} JL * "'V

Fig. 1 Elevation of reactor system

1. Core2. Graphite Reflector3. Sucking Tube4. Primary Cooling System5. Additional Shielding

106

n

Fig 2 Reactor Core arrangement

Fuel clement

Beryllium

Irradiation channel

Pneumatic transferlube

j @J Safety rod

l vOj Shim rod

Regulating re

Neutron trap

107

The reactor control system consists of 2 safety and 4 shim rodsof boron carbide and 1 stainless-steel regulating rod. For the active coreconfiguration of 92 fuel elements and neutron trap the excess reactivityis 10 S e i f while the total reactivity worth of the shim rods is 12 Be£I.

The main reactor parameters are presented in table 1. Themaximum thermal neutron flux in the center of the neutron trap reached

12 22.1 x 10 n/ cm . s ; a value which shows the high quality of the recon-structed reactor. The core cooling mechanism is basically natural con-vection as in the former TRIGA reactor but the driving effect is enhanc-ed by a 2 meters long sucking tube placed directly above the active core.The whole active core-sucking tube assembly is suspended from the topof the reactor tank. In table 1 are presented also the typical heat transferparameters. Due to the large heat transfer surface of the VVR-M2 fuelelement the average heat flux in the active core is rather low as comparedto the TRIGA reactor. The temperature of the fuel element cladding hasnot yet been measurd, but it can be estimated from the parameters givenin table 1 and its maximum value will remain below 95°C.

From the measurement of power versus time after insertion ofan excess reactivity (fig. 3) the negative temperature coefficient of thereactivity can be estimated as:

^ ^ -1 .5 x 10 ' 4 6K/K °CdT

However, this rather high negative temperature coefficient isassociated with the coolant-moderator, while in the case of the formerTRIGA reactor the negative temperature coefficient ( - 1.10" ^ <$ k/k°C)is prompt and associated with the homogeneous fuel-moderator element.As can be seen in Fig. 3 a normal steady state of operation can be safelyestablished with an accidental excess reactivity of ^ 0.5 3 -r

III. SPATIAL AND ENERGY DISTRIBUTIONOF NEUTRON FLUX

The neutron flux and energy spectrum are determined by theuse of detector foils Dy, Au, Lu, Mn, Co, Cu, Al etc. together with agamma spectrometer using HP Ge detector. The experiments are usuallycarried out at a reactor power of 50Kw. The corresponding maximumneutron flux at the center of the neutron trap is obtained as 2.1.10 i 2 n /cm 2 s . The coefficient of non-uniformity in flux distribution ^max/^is 1.24 axially (as measured along the neutron trap axis) and 1.84 radially.

108

TABLE 1

Typical parameters of the 500

Fuel element type

Uranium enrichment

Cold,clean critical loadingUranium

Beryllium

Operational loading

Active core volume

Excess reactivity

Maximum thermal neutron flux

Number of control rodssafetyshim

Peculating

Total reactivitv worthof shim rods

KW Dalat Peactor

WPv-M2

36% U-235

72 elements2.781 Kg U-23526.25 Kg

92 elements3.695 Kg U-235

64.9 1

10 B

2,1.1013 n/cm2s

241

12 8PeffReactivitv worth ofrequlatina rod

Temperature coefficientof reactivity(associatedwith the coolant-moderator)

Primary coolinq systemFlow area of active coreHeat transfer surfaceAverage heat fluxMaximum heat fluxInlet coolant temp.Exit coolant temp.(Average)Coolant mass flowMass flow rate in

primary circuit

0.

-2.10

e f f

Secondary cooling systemMass flow rateHeat exchanger inlet tempHeat exchanger exit temp.

550 cm2

22 m2

2,3 W/cm2

5,2 W/cm2

28 °C50 °C22,300 Kg/h

50,000 Kg/h

90,000 Kg/h19°C24 °C

109

100 400 t(S«c.)

Fig. 3 The power level, as a function of time, achievedwhen the excess reactivity is suddenly inserted.

The measured thermal and epithermal neutron spectra were fitted on theformula:

-E/KTe ( 1 )

Where E is neutron energy, T-temperature of neutron field,A (E/KT)— junction function. The epithermal neutron spectra in theneutron trap and in the active core are presented in Fig. 4. The Maxwel-

110

10'

to1'

to'

10*lft *<«V»

Fig. 4 Epithermal neutron spectra in the neutron trap (o),active core (o) and rotary specimen rack (x).

lian part of formula (1) can be deduced after having determined tempera-ture T (using Lu foil). If the effective neutron temperature can be ex-pressed as function of the temperature of the coolant-moderator TQ

according to:

T = a T o

then the coefficient ex can have the value 1,08 for the active core. Thevalues for parameter $ are 0,03; 0,02 and 0,10 corresponding to theneutron trap, graphite reflector and active core.

The results obtained as reported dn this part are currently usedfor the evaluation of possibilities for different activation analysis purposes.

111

IV. ACTIVITY OF THE PRIMARY LOOP WATER

The former Dalat TRIGA MARK-II reactor became operationalin 1963 but in reality it has been kept idle for many years before itsreconstruction- As a consequence, many old components in the reactortank are suspectedly under the attack of surface corrosion. In addition,the pollution of the reactor components in the tank during the processof installing cannot be removed completely. In these circumstances,although the primary loop water was carefully treated and its maincharacteristics (PH, conductivity etc.) kept at the nominal values, thecontent of various radionuclides in water must be accurately determined.In table 2 are given the specific activities of various radionuclides inprimary loop water after a 72 h of continous operation of the reactorat 500 Kw. Most of the radionuclides presented in table 2. resulted fromthe activation of the corrosion products of aluminum alloy and stainless-steel. The presence of Xe-135 indicate the probable release of the noblegas fission product, through the microdefeats of the fuel elementcladding.

TABLE 2 .Specific Activity of radionuclides in primary loop water after

a 72-h continous operation of reactor at 50GKW

f

Hadio-nuclide

28A12 7 r , g

Na56Vn51Cr

w3 8ci

Ar

Xe

Ti

2.25 in

9.46 m

15.00 h

2.53 h

27.7 d

. 23.8 h

37.18 m

1 .83 h

9.17 h

Act iv i tv(Ci/1)

U6.10" 7

5.10"6

5.10"7

1 .5.10""7

1.10""8

1.10"°

1.10"7

2.10"6

2.10~9

112

INTERNATIONAL ATOMIC ENERGY AGENCY

M SEMINAR ON APPLIED RESEARCH AND SERVICE ACTIVITIESPOR RESEARCH REACTOR OPERATIONS

CoDenhapen, Donmark, 9-13 September 1985

IAEA-SR-119/20

NEUTRON SPECTRA OF THE NEWLY RECONSTRUCTED

500 kW NUCLEAR REACTOR OF DAWT, VIETNAM

APPLICATION IN ACTIVATION ANALYSIS

PhcMTt Zuy Hien ,Tran Khanh Mai ,Ngo Ouanq Huy ,

IJriuyen Tac Ar.h

D a l a t N u c l e a r P c s e a r c h I n s t i t u t e ,Da l .at , Viet.nair.

'ULJLI H A c T

The neutron .spectra in different irrar.iat.iw!: positions in the r.ewly

reconstructed 500kW nuclear research reactor of Oalat,Vietnam,havc been

measvired.The results obtained are used for evaluating the advantages

of each irradiation position in activation analysis and for the absolute

determination of the concentration of elements through t.he ko-standardi:»a-

ti.on technique.More than three thousands samples have been analysed and

the results show the high performance of the newly reconstructed reactor.

This is ,i preprint of a paper intended for presentation at a scientific nu»M«nrj. Because ol The provisional naioro c' 'Ucontent and since changes of substance or detail may have to be made before publication, the preprint i:. < i.:ic available on t*iounderstanding th-it it w i l ' riot b« cited in tlic literalurp or in any way bf reprorjucod in its proscni f e r n . Tru- v«ow<- oxprcs.f'J ^nth.' suM''fT»*;iit'. n>dd»; rern.T-n th<? luspoiri ibil ity ol tho named iji.-Uioi (s). the vii-wr. do no: n»i.cssiiily reflect thorvc c ' ths novi;.*r<-nior.i ol 11>f ilcsifinatifiQ Mv i t i t t r 3loir'.s/ o< ol the dosi^nasinp orcjjir.ijiitionlsi /•• o\nicu!.T "Either .'/)<•• IAFA no< nny t>theru f ' i . ' ; i i i 7 . i ( m t i i n i " i l\ : p o r , s ' i n r , g I ' m r n r e n n y Ci'H t > r ' • t ' l r J r u s p o m i l ' l " • ' " ' J ' < y < i , > ' - r i . i ' ' r a r . t d r c r - l . n l , ' ; , s ,•>, • • p r i n t .

113

! n i i oiiuct" ion :

The newly r e c o n s t r u c t e d 500kW n u c l e a r r e s e r n c h r e a c t o r of Da l o t , b r o u g h t

t o i t s n o m i n a l opordf. im] power i n F e b r u a r y l r > R 4 , i s u n i q u e of i t s k i r.r, i r t h e

w.-ir1 d :Sov i .M . i e s i q n e d c o r o nnd c o n t r o l s y s t e m h a r m o n i o u s l y i ns r..-1 ! ••'! in t h e»

l o f t - o v e r i n f r a s t r u c t u r e of t h e o l d A m e r i c a n made 2 5Ok"W TRIGA MARK-:'! r n n c t . o r

"he u n d i . s n i o a n t n b l e c o m p o n e n t s o f t h e f o r m e r r e a c t o r . s u c h a s the? r c - f L o r t a n k

and c o n c r e t e T . h i e l d i n q , t h e g r a p h i t e r e f l e c t o r , t h e beam t u b e s and L!,' : m a l c o -

lumn a r e :•••> ':>•.• r e t a i n e d ,wh i l e nQ.w d e s i g n work h a s b e e n c a r r i e d o u t : o r s u c h

n'->w c o m p o n e n t s .ind s y s t e m s a s t h e r e a r t o r c o r e u s i n g S o v i e t f u e l '• i t:n<*n ry

W P -M'2 ,!;hf '.-'^r :•. r ^ l s y s t e m , t h e c o o l i n g syr i tom and t h e w h o l e r- ic i io- i . ro" ' : ( . : l i o n

.•md s a f e l y ::vsr.»;-ins of t h e n u c l e a r c e n t e r . Tin: fo rmer b u l k ,sh i ^ ld. inu -tank h a s

b(?-j:i r e c c n - ; ! r u c f . v j i n t o o s t o r a g e t a n k j ' o r b u r n t - u p i ue 1 o] o r i e n t s .

I_I_. \;o(rv? f undarr-.er.t a l c h a r a c t e r i ^ 11.cs o f t h e r t r c n y t r r . c t c d r e .• ':'.>r :

The : MI 1 " l e n i e n t VVR-M2 i s inaile of ur-nn i um-a ! urni p. i urn a l l o y wifl^. iCi\ U

r n r •• r-hmen t . I ;; c o n s i s t s of t h r e e c o a x i a l t u b e s w i t h thv t - .hickr .ess o1 . : . r>mni m-:

rilic.'wn in ! u i . : . T h e c e r e <:r<n f i q u r a t i o i i i s p r e s e n t e d i n : iy. 2 . S e v e n - .r . i r c o l l s

in r h e ' . v r i t p r of r h e r o r e a r e o c c u p i e d by a w a t e r - b e r y i i urn n e u t r o n ;.r Lip . w h i c h

l ias i n s i d e d i ame to r ' ^Smm . The w a l l o f Mir n e u t r o n t r a p i r. marie of be? y 1 i "in . Thn

r o n i o n r e t w e e n t h e f u e l e l e m e n t p a r t of t h e c o r e a n d t h e o l d g r a p h i t e r e f l e c t o r

i ? a 1 r;o l i n e d w i t h b e r y l i u m . T h e r e a c t o r c o n t r o l s y s t e m c o n s i s t s of 'j s a f e t y >\v\A

'! s i i i i i r o d s of h o r o n r v i r b i d e ^ l s t f ai n l o s s - s t e o l r e g u l f l t i n q ro-:. . F o r ; !i;: c o r e lo<a.-

dincj of 9? f u e l e l e m e n t s t h e e x c e s s r e a c t i v i t v i s 10 Q . . . . w h i l e t h e r e a c t i v i t y

w o r t h of e a c h s a f e t y and sh im r o d i s J p . _ .j e f f

The c o r e c o o l i n g m e c h a n i s m i s b a s i c a l l y n a t u r a l c o n v e c t i o n a s i \ t h e f o r -

mer TRIGA r e a c t o r , b u t ' .he d r i v i n g e f f e c t i s e n h a n c e d by a 2 - m o t e r lorv.) cuckintf"

t-'ibo p l a c e d d i t a r t l y a b o v e t h e c o r o . The; w h o l e c o r c - s u c k i n q tube; as'-."rr: l y i s r^U5-

i e n d e d from th<- t.r-p of t h e r e a c t o r t a n k ( f i q . 1) . Due to t h e I a r q n h i - \ ! t r a n s f e r

•"••lvface o f i he VVU-M.? fue l e l e m e n t t h e a v c r a q e h e a t f l u x in t h e <-or.- i s r a t h e r

•ow .is com| )a r ' - i t o Mie f o r m e r TRIGA r e a c t o r .The h e a t t r a n s fr-t . jv l i r / d i o u l i c p 4 . -

r \\v r i-vr, r>'~ H v ' • ' " • o r ; ! r ,"-»;ed r e a c t o r ^n^ q i v n n i n Tal-j]o 1.

' x p e r i m c i i t . i 1 ."::•> - a ; '••u ! a t i o n a l w o r k s ; r e in p r o u r t 'S : ; in -srder i • i n * r e s t i -

I 11 r- :n i nt er r %st i na .irui i a i f < i r t a n t p r o b I em ; w h e t h e r f*hi: norn ina! power -if t.!-.e :i«'V

[ ' • .K ' t r i (Ticiy br- u | ' ; r . u ! - ' l t o more t h a n 5<)f,'kW w i t h o u t s i cm i f i r jp. t chav. ' ie in t h e

' • < > : i r , ; i I K - C \OT\ iM' . t l r o . i i - r f i r c o m j i o n e n t s .

1 1 1 . _ . 1 f : i t i a t a r n i e n e r g y d i s t r i b u t i o t i n ^ t h e n e u t r o _ n _ f l u x :

T h e r . e i i 1 - r ^ - n f l u x a n d e n e r g y s p e c t r u m «•»-<-> d e t e t m i n e d b y t h e u s - ' ••: d e t e c t o r

t o i l s D y , A i t , L u . M r i . C u . C o , T n , W , A l ( f o r t h e r m a l . i n d e f i i t l i e r m a l n o u f r o n s ) a n d U n , . 5 , *

r1-. / ' o , p.h , A 1 f f - > r f a s t n o u ' t o n s ) . - T h e a c t i v a t t d f o i l s ^ r o m e - i s u r e d o n M i ? q a m -

114

1.5r- 10.5 oj 03

Al Al

Fig. 1 Cross section of the fuel element( Unit in mm )

Cell 1-1

/

Fifi 2 Reactor core arrangementI. Fuel element ; 2. Beryllium reflector 3. Neutrontrap ; 4. Wet channel ; 5. Pneumatic transfer tube ;6. Safety rod ; 7. Shim rod ; 8. Regulating rod ;9. Graphite reflector ; Cell 1-1 : The first cell of thecore, cell i-j denotes the cell at i t n row and j t n column.

115

.1 r* . 3 Vrrtif-nl soct ior. of the roconsf met'"-! reactor

1 . • : , ) - . • ' •

2. C,ni\)h i l.c He T loo tor

j . Duel; !. nr; T u b e

4 . V r i r:i't>-y C o o l i.nr. -r>v" '• rn

'•'•. .••.'!•! • ' io'Ki.1 . I h i o ' l ' l i •:.•;.

116

n>;i «;j-oc: •"•"•nv.'tor I'.sii.j !IP(> d e t e c t o r . The maximum thermal ncuLior. 1*1 ux a t t h e1 \ 2

eon tor nf the neutron trap is 2.14xl0 n/cm .s.The coefficient of non-unifor-

mity in flux distribution is:7 max .

—u.~ " = ^ • rad i <il 1 y ,and

i)Vi I .

- 1.24 axially (as measured along the neutron trap

Tiiv flux di .str i bui ion in the neutron trap has been measured \n detail

in order to investigate the possibility of using irradiation channel for neu-

tron df'ti.nj in silicon.The results shov that the flux uniformity of + '>* is

ohstjrvuc! in the :euion <-.f 5cm in diami.'i.or «ind i'.Qcvn in lieiqht around the cen-

ter oT t he neul-.ron t. rn|>.

The measured thermal and epi thermal ncu'tron spectra were fitted on the/1 /

well-known formuia' :

A (F/kT)( 1 ! •

wy.r-ro: F - noution enerqy,

T - effective neutron t<D!r,i>erature,

£ (V./kT) - iunction function. .

Tim r?ffective neutron temperature is measured by using a non- — Lu-de-

tector .T!i'- calibration experiment has been done by irradiating the Au and Lu

foil:- in thermal column,whore the ef f r;rt: iv<? neutron temperature is obviously

rq-jol ic t lc temperature of graphite! 3^0 K at 500kW) .

The- 'svrameters (£> and O^ are determined bv irradiation of Cd-coveredr e\ni os'••!', ir><v: Jet.eivtors. In t.his case the reaction rate per nucleus of the irradia

/ 2 /t <'<l i.'ii '-.t,i bf£ ralculnted by u.vincj the formula ' :

A - <b . . 1 UTC ) . F ^ , , ( 2 )' epj o Cd

wh« ••.:• I r, . ; s the epi t hoTna.l neutron t ransmission fac to r for Cd-f i l tr: r (F ^"1)

.in-i : i<y. ! ; •: ;. ho rusonaiiCfi in teg ra l modified by the devia t ion of t ho e p i t h e r1

ni.i ! noiif rc.\ •;; ».ctnra from th.e — - lav:I - 0 . 1 2 ' ) ^ O./i:-"^

«> o ' o

(f: ) * O.L.Sx(2uc + 1} '

v h r - " : : 1 - ' . -onv^nt i o n a l rf?r.--.>nance i n t o n r a l ( oC - 0) ,

0s- - ' i i 'Tn ia l •!• -«i I »• «• »n ix\ ,V) c : r o : ; s - : ; e c t i<in ,. / 2

F - - ! f ( ^ c t i v f > ••son.-mce e n e r g y (eV) , a s d e f i n e d ond t a b u l a t e d i n Ref

In T,u;l" .' .ii' | • •;'..-i:t.ed the oxpi i imental values of the neutron r.pectra

i " ' - - i r . \>, ! f > i : r t y p i - - . i l i r r a d i a t i o n s i t e s . T h e p a r a m e t e r (y , / y> . i n c r e a s e s

i ' i , r . i : ' l •• f t i .m t I K i r i n \ i n L i o n c h . i n n e 1 ( 1 ' - ? ) i n t h e c o r e t o t h < - r o t a r y

':•>••:, c • i i r M i ' - M i . i i l M t i r e M i " - - ! m . I n t h o i l i e r m . i l c o l u m n CD . ^ <P . , - 3 P . « i

117 pOOROUAUTYlmsss

therefore the epi thermal neutron spectrum c.-jn Ln.- ti'. i I'• r o.| j n ,K:T. ivul ion

technique.The deviation of the epithcmrmal lunit.roti sp'.-vmun from \);r ideal

— - law i<; evident ,especially for the irradiation chanre If I )-2) .Ar; .--HI b*-.:

seen later,it is very important to take into accMini '.iis i; i. t.iMt i.-n in the

mil' t.i element activation analysis.

Table 1

Typical parameters of the 500kW_Dalat roac.^tor .

Ku»?l element type

Uranium enrichment

Cold,clean critical loading Uranium

Beryllium

Operational loading

Active core volume

Fxcess reactivity

Maximum thermal neutron flux

Number of control rods

Total reactivity worth of i:h j.m rods

inactivity worth of regulating rod

Temperature coefficient of reactivity

(associated with the coolant-moderator)

Primary coolinq system

Heat transfer surface

Inlet coolant temperature

Exit coolant temperature(average)

Mass flow rate in primary circuit

Secondary cooling system

Mass flow rate

Heat exchager inlet temperature

Hoat exchanger «jxi.t temperatuer

WR-H;:

72 elemc-nts ,-3-r7-ffl iA) I.i

2.:>.;?5kg

'''•1.91

10.

2.1x10 'n/c/.s

7

12 B ..

J e r r

50°C

'30,00('ka/h

90 #000kg/h

19°C

M°C

Table_2

Tyj'ical charactor i? t jc .s neutron .spect ra in some irradi<.!• ion sites_( .'*: r>0kW) .

Neutron t rap Core Ro^iry Specimen Thermal column

(Uni t c e l l ( 1 1-?)) Rark

th(n/em's)

T(°K)upi

4.2X1O1

25.3

331

-0.1 1

100

310

1.5x10

1574

300

I V . / > I I I i r . i t i r > n i n . A t i v a t . i o n A t 1 1 " . •". i •; :

118

The i ri-cKli.-it. ion f a c i l i t i e s for o::\ i vul ion .m.ilyiiis <U the Mixleiir n»-

sernch reactor of Da]/it consist of,:

- one neutron tra:> at the center of the core;

- a pneumatic transfer system ins ta l led at. the perimeter of the core

{unit ce]i (13-2)) ;

- a rotary specimen rack with 40 i r radia t ion inmitions in the. graphite

ref lec tor ;

- an automatic pneumatic t ransfer system ins t a l l ed in the thermal co-

lunn for trherm-nl neutron activation analysis of shor t - l ived i so topes .

A neutron counting unit with six 'lie-neutron counters is also coupled

to this; sy«;tom for analysis of uranium by del ayed-nc-utron technique.

Tn Hie ricvir future two automatic pneumat i c system will be i r»s< a 1 led ,one

ii, the cor" on;: the ol.hc-r in one of the horizontal tub*-:; .

The experimental resul ts given in Table 2 have, boi-ri used to evaluate the

a -Ivan Lagos of each i r radia t ion s i t e s for different analyt ical tasks;.

For t!iis purpose the s ens i t i v i t y of act ivat ion analysis of various e l e -

mrn's has l:.-v;ri calculated for different i r rad ia t ion pesi t ions .The r e su l t s she

tli.it-nrat.h--M 3 ,-»rq«> variation in the absolute and r e l a t i ve sens i t iv j Lies can be

found by chcinijing the posit ions and conditions of i r r ad i a t ion (wi ttv.ir and witVu

Cd <-)ver) .This ccii-\ he seen from the example of INAA of qold in ores con-

taining antimony as a m.-ijor component. In th is case the presence of the very122 2 2*1

intensive phot ope akr of Sb " and Sb (!>64keV and 604k«V) in the grirona spec-

trum of the irradiated sample creates a high Compton backqround and disturbs198

the determination of the 412keV photopeak area of Au .It is shown that if the-irradiation is carried out in the rotary specimen rack instead of in unit, celt

(13-2) one can reduce about 40% the relative activity of both antimony isotope*198

with respect to Au .The important role of the condition of irradiation can

he demonstrated in activation analysis of uranium in the ores containing rare139 141

(>arth elements. Tn this case the interference of the isotopes Ba ,O and140

I.a can b«» as» 50 t imes r educed i f t he sample i s i r r a d i a t e d in t h e r o t a r y s p e -

cimen rack under Ql c o v e r .

^ n ^ t h r i ry.ami.le i r t he a c t i v a t i o n a n a l y s i s of nrar. riim by I rr<sr\ \ ;>!". inq t h e

ymip i r in !. he tnt-rmril lolumn and cot int i ivj t h e ilfilayud r : eu t r o n s . Al t:t«)uyh t h e

sori.T i t i vi tv in ' lit r, enso i s n^ t r.o *high(.it t h e pjwi l o v e ! )r!i!P t o Ihv r a t h e r lov;

tilieim.'il noL.rr<.iii ! 1 MX in t he thermal col utnn ,thc; i n t o r f o r r - n c o from thot ium can be.

c<"»mpl r>t r 1 y nncj l/..•»: i i_'d .Th i s s i t u . i t j o n i s v«ry important : w'o- n dot .ormininq t h e u r a -

niiur. r:ont(>nv. ;. u monaz i t c s a n d , w h e r e t h e t h o r i u m con t en t i s of some o r d e r s of

n- r . i : i » r u d e h i n i i ' . - r U w i n i ; r . i n i u m .

'Vhc : i r i r . v n r ! - t •«, o f t h e . r e a c t o r n e u t r o n s p o c ^ r . i <i i v e r i n T a b l e 2 . i r e a l s o

!• - - d f o r l b s o i Hi . - d n t t . M T T i i n . i t i o n - i n m u 11 i e I f i n e n t a l i c t i v . i t - i o n , - j n a l y s i s t h r o u g h

\: - i ^ i . ' i n d T r d i ?.<-»t i o n t t ' r h n i « i u e . C o ) d h a s b e e n ' h o s " n . i s s t . i : i d c J i < l a n d tJn .» V - d a t aii O

119

n r o t i k o n f r o m F<il . ' . A n - l n o n c e m n l r t i . i l ••• i I h k n o w n c o n c e n t r a t i o n o f d e m o n t

hi t s brvn <"m,» I y :•;•."] t n p i o v o t h o w o r t h of a p p l i c a t i o n o l t h e k " S t a n d a r d l y a t i o no

tv-chn i ' t u p .

I t ?h.)i i lr l !,•• nif?nt icm.><! r h . i t v e r y h i - jh o c c u r . n c y i r; o b t a i n e d v/hnti \ i s i t ; g the

k . - s t a ru ' lu r r l i z a t i On t »;r hn i<pm i n t h e n n o l nrvit r o n " s r t i v a t i o n n n < i l y s i s . T n t h i s c

s e t h f a c c u f . ' i c y f t t-ho C X J X T i m e n t a l r e s u l t s i s p r a c t i c a l l y n o t n f f («:t.f>rl by thtf

o r r o r r . o f thin rrn».v.nrr(i n o u t ron" . s p e c t IVI ;vit . l t n o t o r s .Th i s i s c o n f i r m e d by t h e e x -

p o r LniMits c . i r r i f - . ) <>nl in ' ihr r.hrtrmol c

nefo fences

1 / K . H . B e c k u r t s ,K.Wir tz Neut ron = P h y s i c s S p r i n g e r V e r l a g 1964

2/F.De c o r t e ««• a l . J . Radio nna.1 .Chcm. 6 2 , ( 1 9 8 1 ) ,209

3 /A. Ahmad e t a l . J . R . id ioana l . Chem . 72 , (19U2') ,335 .

120

Proceedings of the 9th Pacific Basin Nuclear ConferenceSydney, Australia, 1 -6 May 1994, Volume 2, pp. 1033 -1040.

VN9700005

INVESTIGATION OF THE BASICREACTOR PHYSICS CHARACTERISTICS

OF THE DALAT NUCLEAR RESEARCH REACTOR

NGO QUANG HUYCenter of Nuclear Techniques, Hochiminh city, Vietnam

HA VAN THONG and NGO PHU KHANGNuclear Research Institute, Dalat, Vietnam

SUMMARY The Dalat nuclear research reactor was reconstructed from the TRIGA Mark II reactor, built in 1963with a nominal power of 250 kW, and reached its planned nominal power of 500 kW for the first time in February1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of itscharacteristics is carried out by the determination of the reactor physics parameters. This paper represents theexperimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source leftafter the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radialdistributions of thermal neutrons, the safe positive reactivity inserted into the reactor at a deep subcritical state, thereactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc.

1 INTRODUCTION.

The Dalat nuclear research reactor was reconstructedfrom the TRIGA Mark II reactor built in 1963 with anominal power of 250 kW. In 1975 all the fuelelements of the TRIGA reactor were dismounted andtaken away, which effectively caused the reactor to stopits operation. Reconstruction of the reactor with a highoperating power of 500 kW was accomplished with aRussian technological assistance. The new reactorreached its planned nominal power of 500 kW for thefirst time in February 1984.

The Dalat reactor is unique of its kind in the world :The Russian designed core is harmoniously integratedinto the left-over infrastructure of the American-madeTRIGA research reactor. The undismountedcomponents of the former reactor are the reactor tank,the concrete shielding, the graphite reflector, the beamtubes and the thermal column The new componentsconsist of the reactor core, the cooling system and thereactor control system. The rector core is loaded withthe Russian-made WR-M2 fuel elements of uranium-aluminum alloy An amount of beryllium material isplaced at the center of the core, forming a neutron trap,and at the periphery of the core, forming asupplementary reflector.

The Dalat reactor has three characteristics distinct fromthe former TRIGA reactor :

i. The WR-M2 fuel elements of uranium-aluminumalloy have been substituted for the former TRIGA U-ZrHn fuel elements. It means that the low inherentlysafe reactor core took place instead of the highinherently safe one.

ii. The introduction of a neutron trap in the center ofthe core and an additional beryllium reflector beside theold graphite reflector enhances the neutron flux at thecenter and makes the neutron field distribution morecomplicated.

iii. A delayed photoneutron source induced by theinteraction of gamma rays of energy greater than 1.66MeV on beryllium material may have influence on thereactor's kinetic behavior and the reactor operation.

Investigation of the characteristics of the Dalat reactoris carried out by the determination of the reactorphysics parameters The experimental results wereobtained for the effective fraction of the delayedphotoneutrons, the extraneous neutron source left afterthe reactor is shut down, the lowest power levels ofreactor critical states, the relative axial and radial

121

distributions of thermal neutrons, the safe positivereactivity inserted into the reactor at a deep subcriticalstate, the reactivity temperature coefficient of water, thetemperature on the surface of the fuel elements, etc.

2 REACTOR DESCRIPTION

A view of the Dalat reactor is presented in Fig 1, inwhich the new components consist of the reactor core,the extracting well and the cooling system. The reactorcore in the cylindrical shape of 43.6 cm diameter (Fig2) is surrounded by the old graphite reflector of 34.6cm thickness. The reactor is loaded with 89 fuelelements of WR-M2 type. A fuel element has anoverall height of 86.5 cm including 60 cm uranium-

11 MFig. 1 Elevation of reactor system1. Core; 2. Graphite reflector; 3. Extracting well4. Primary cooling systeem ; 5. Additional shielding

Cell 1-1

Fig. 2 Reactor core arrangementI. Fuel element ; 2. Beryllium reflector 3. Neutrontrap ; 4. Wet channel ; 5. Pneumatic transfer tube ;6. Safety rod ; 7. Shim rod ; 8. Regulating rod ;9 Graphite reflector ; Cell l-l : The first cell of thecore, cell i-j denotes the cell at i t n row and j " 1 column.

aluminum ftiel part (36% enriched U ^ 5 ) that consistsof three coaxial tubes of 2.5 mm thickness. A tube ismade of a 0.7 mm fuel layer wrapped in two aJuminumcladding layers of 0.9 mm thickness. The average massof U ^ ^ i n each fresh fuel element is 40.3 g.

Irradiation facilities inside the reactor core consist of aneutron trap, a wet channel and two vertical pneumatictransfer tubes. The reactor is controlled by six boroncarbide rods (two safety rods AZj, AZ2 and four shimrods KC1-T-KC4) and an automatic regulating stainlesssteel rod AR. An additional beryllium reflector layer of2.9 cm average thickness is inserted between the activecore (the fuel region) and the graphite reflector . Theequivalent radius of the active core is 18.9 cm.

3 DELAYED PHOTONEUTRONS

Delayed neutrons in the Dalat reactor come from thetwo contributions : the delayed neutrons from U2-^fission and the delayed photoneutrons from (y,n)reactions induced by the interaction of gamma rays ofenergy greater than 1.66 MeV on beryllium material.Determination of the delayed photoneutrons plays animportant role in studying the reactor kinetic propertyand estimating the neutron source left after the reactoris shut down.

The time spectrum of the delayed neutrons is measuredwhen the reactor is shut down after continuously 100hours operation at 500 kW (Fig 3). Fig 3 shows that thedelayed I P - " fission neutron counts decrease rapidlyand they may be neglected from 5 t n minute afterreactor shut-down. The delayed photoneutron countsdecrease slowly and remain about 10"^ -f- 10"^ of thenominal power during 64 hours. So the photoneutronsource left after reactor shut-down gives the neutronflux of about (105-r 108) n/cm2/s during 64 hours. Thisextraneous neutron source is used for starting-up thereactor.

JO 40 50 t(min.)

Fig. 3 Time spectrum of the delayed neutrons afterreactor shut - down.

0 ExperimentFitting curve for 6 fission nutron components

and 10 photoneutron components- - - - Calculation curve for 6 fission neutron

components

122

The delayed neutron spectrum is analyzed in 16components comprising of 6 components of delayedl £ 3 * fission neutrons and 10 photoneutroncomponents. The total effective fraction of the delayedneutrons Pefj consists then of two parts: P eff»effective fraction of the delayed fission neutrons, andP®e

eff, effective fraction of the delayed photoneutrons .

A global estimation from the time spectrum gives :

PBeeff = ( 0.49 ± 0 . 0 2 ) % p e f f

P U eff= «- PBeefT= 99.51 % p e f f .

Table 1 presents the P eff'Peff r a t ' ° a s a ^ u n c t ' o n °fthe nige/ni|j mass ratio for several various media. It isobvious that the P eff'Peff r a t ' ° ' n c r e a s e s with thenige/my mass ratio .

Table 1. Pmass ratio.

r a t ' ° a unct'on

m Be / m U

8.51 7 7 7 - 3112

Pure beryllium

PBeer#eff

0.49 %2.26%2.31 %

Reference

This workKrasin, et al, (2)

Keepin, (1)

4 LOWEST CRITICAL POWER LEVELS

The determination of lowest critical power levels of theDalat reactor containing an extraneous source isinteresting from the safety view-point in the reactorstart-up process and from the requirement of findingpossible low powers for the reactor experiments.

The experiment is carried out at 10ln day after thereactor is shut down followed continuously 100 hoursoperation at 500 kW. The two safety rods arewithdrawn completely, the four shim rods are placed at29.5 cm in depth and the regulating rod at 35 cm. Thereactor power is 6.7 10"6 Pn, where Pn is 500 kWnominal power. The reactor power is increased byshooting a cadmium sample from the reactor core usinga pneumatic transfer system. It is equivalent to aninsertion of a positive reactivity of p = 4.67 % pefl-.Neutron densities in subcritical states increase andattain the stable values They are well described by theiiibcritical reactor kinetics theory(Loren C Schmid,(3)).

The critical state of the reactor is observed when therour shim rods are located at 28 4 cm in depth and the'egulating rod is withdrawn continuously from thedepth of 40 cm to the depth of 20 cm. The lowest:ritical power depends on the beginning depth of theegulating rod The values obtained are from 1.6 10"^

P n to 4.2 10"5 P n for the beginning d<t)Tths from 40 cmto 60 cm of the regulating rod.

The range of the lowest critical po\wiR depends againon the time after the reactor is shut down (Fig 4). Theyhave the values from (0.5 -M.2) 10~4 Pn , i.e. 25 W •*•60 W, to (1.1 -s- 1.6) 10"5 P n , i.e. 5.5 w ' - r 8 W, from4 t n day to 13 tn day after reactor shutdown.

(V*tt>

!O

Time after reactor chut - down (day)

Fig. 4 Lowest critical power level* ns functions oftime after the reactor is shut down

£ ExprimentCalculation value of P c so thai ihe first term on

the right hand of the formula (2) equals 300% (a),100% (b), 10% (c) and 5% (d) of the <s*,cond one during1000 sec.

In the just critical situation, k = 1, tli*. neutron densityor power level will rise at a linear rate, as shown byLoren C Schmid (3),.

P(t) - t + Pr (1)

6

m

where P s « (1.93 ± 0.01) 10"8 t>n j s extraneoussource power, P c is critical power w^hout extraneoussource, 1 is prompt neutron lifetime, p ancj \ a r e

delayed neutron fraction and decay constant for the i t n

precursor.

10 days after reactor shut-down the rector power canbe expressed as follows :

P(t) = 5 10" 8P n t + Pc (2)

This formula shows that during the fi/ti |000 sec in thecritical state the first term on the right hand is greaterthan 300% of the second one, which K about 1.6 10"^Pn. It means that the reactor power n not just constantbut rapidly increases . In order to dtr.rease the powervariation to the value of less than 10% \\ j s necessary totake critical power of 5 10"^ Pn .

123

In Fig 4 the critical power is plotted against the timeafter the reactor is shut down so that the first term onthe right hand of formula (2) equals 300% (curve a),100% (curve b), 10% (curve c) and 5% (curve d) of thesecond one during 1000 sec. From Fig 4 it is obviousthat the lowest critical power measured varies in therange of (100 -r 300)% during 1000 sec. Thesevariation limits are too large for practical use of thereactor. They have to be reduced to the values of about(5 -HO) % to sustain the stable criticality of thereactor. Therefore critical power must be chosengreater than 10"3 Pn or 500 W.

5 . SPATIAL DISTRIBUTION OF THETHERMAL NEUTRON FIELD

The replacement of the former TRIGA reactor core bythe new core with the neutron trap at center and theberyllium additional reflector layer at periphery makesthe active core in a rim shape with 4.3 cm inner radius,18.9 cm outer radius and 60 cm height . So thethickness of the active core is 14.6 cm that is equal to 4-r-5 fuel element sizes. The active core is surrounded bythe beryllium and graphite reflectors. The materialabove and under the active core consists of aluminumand water. This narrow active core contains 89 fuelelements, 7 control rods and 3 irradiation facilities.Therefore the spatial distribution of the neutron fieldwill be complicated.

Measurement of thermal neutron field distributions isusually carried out at few watts power by activation ofthe foils put on the fuel element surface. It is difficult toapply this method for the Dalat reactor because of high

3

"3

e0 . 5

1 . 0

0 . 5

1 . 0

0 . 5

b

f•

r

•

t

r

1

•

I

¥

1

*

Cell

* ' *

Cell

\ h4

Cell

a

*•

7-3

' s7-9 ,

*1-4

•

*°

10 5O B(«>)

Fig. 5 Axial distributions of the thermal neeutrons atthe cells 7 - 3, 7 - 9 and I - 4 with different depths ofcontrol rods

co3<U

Z

i.o

0 .8

0,6

critical power as mentioned in Section 4. Therefor*measurement is carried out by foil activation in watecolumns or on surface of a dummy element made oplexiglass instead of fuel element. In the experimenthe Cu foils are irradiated at 2.5 kW in 15 minutes.

In Fig 5 are presented the axial distributions of tlvthermal neutrons measured in water columns at cells 73, 7-9 and 1-4 for some combinations of the controrod's depths. Fig 5 shows that almost of the curves anin dissymmetrical shapes which can be expressecanalytically by Tong and Weiman (4) as follows :

d>(z) = [ A + C(z-30) ] cosBz(z-30) (3)

where A, C, B z are constants, C = 0 for symmetricacurve and B z is the axial buckling.

The radial distribution of the thermal neutron;measured on surface of a plexiglass rod placed at cell:4-1, 4-10, 5-2, 5-3, 5-10, 6-3, 6-4, 6-5, 6-8, 6-9, 6-10 i;shown in Fig 6. It is obvious that the curve hamaximum value at close by the neutron trap, decrease:rapidly within the first fuel element, after thadecreases slower and finally raises at close by tinberyllium reflector. This distributiort can be expressecby the two-group neutron diffusion theory for a banreactor , as shown by Murray (5) :

<D(r) = a] J0[a2(r-a3)] + a4 I0[a5(r-a3)] (4

where JQ and IQ are Bessel functions, a\+a$ anconstants.

The constants in expressions (3) and (4) are determinecfrom experimental data by the least square fitting

At the center of the core (R = 0)O = 3 36 ± 0.04

16 IB R (<

Fig. 6 Radial distribution of the thermal neutrons iithe reactor core

o ExperimentCalculation , expression (4)Calculation , function aj Jofa2^r" a3)J ' n C4)

124

method So geometrical parameters of the core such asinhomogeneous coefficients of neutron field in axialand radial directions, axial and radial buddings,effective height and effective radius, axial and radialextrapotated distances are obtained (Table 2). In Table2 are also presented the calculation values ( Technicaldesign (6) and Nguyen Phuoc Lan , et a l , (7)) of theabove-mentioned parameters which are well agreedwith the experimental results.

Table 2 Geometrical parameters of the core

ParametersInhomogeneous

coefficient

Buckling, m"^

Effective size,cm

Extrapolateddistance, cm

Radial1.77 ± 0.11

1.71(a);1.74(b)

84.6 ± 5.584,6 (b)

27.6 ± 1.026.2 (b)

8.7 ± 1.58.8 (b)

Axial1.32 ± 0.01

1.35 (a,b)

16.6 ± 0.616.4 (b)

77.0 ± 0.977.7 (b)

8.5 ± 0.58.8 (b)

Note : (a) Calculation result from Technical design,(6)(b) Calculation result from Nguyen Phuoc Lan,

et al, (7).

6 TEMPERATUREREACTIVITY

COEFFICIENT OF

For the light water moderated reactor, temperaturecoefficient of reactivity is defined as the ratio aqp =dp / dT. where dp is reactivity variation correspondingto the water temperature variation dT in the core.

For the Dalat reactor OLJ is measured at 0.25 kW in thetemperature range of (20 -r-40)°C by compensationmethod using the regulating rod. The result is asfollows:

a T = - (0 8 ± 0.1) 10"4/°C for T = (20 -f- 30)°C

a T = - (1.2 ± 0.1) 10"4/°C for T = (30 -4- 40)°C

This experimental value is not in contradiction with thedesigned value a j = -1 10"4/°C (Technical design(6)), and falls within the range of reactivity temperaturecoefficient for research reactors moderated and cooledby water, as indicated in Directories of NuclearReactors (8).

Although a-j is relatively large for water moderator, itstemperature effect does take place with some delay,whereas reactivity coefficient due to variation oftemperature inside fuel elements gives rise to a moreinstantaneous effect. For WR-M2 type fuel of theDalat reactor, as shown by Technical design (6), thisvalue is -0 02 I0"4/°C , while for TRIGA type fuel,

Technical Foundation of TRIGA (9) shows that itequals -1.3 10'4/°C. Therefore TRIGA reactors possessintrinsic stability much greater than the present reactor.

7 POSITIVEBEHAVIOR

REACTIVITY INSERTION

Nuclear reactor is a controlled system with feedback.From the dynamics point of view, it is then importantto study stability property of the reactor undermodification of dynamic parameters.

For its behavior studied during a short time after arapid insertion of positive reactivity, Xenon poisoningeffect can be neglected, reactor stability is then assuredonly by negative feedback of moderator and fueltemperature effects.For the Dalat reactor, stability has been studied at 0.1kW for different values of reactivity inserted bywithdrawing control rods. A positive reactivityinsertion makes reactor power to increase versus timeand after about 10 minutes to attain the stable powerlevel due to negative feedback of water and fueltemperature effects. The dependence of attained stablepower levels on inserted positive reactivity is shown inFig 7.

P(kW)600

500

400

300

200

100

0.1 0.2 0.3 0.4 0.5P (Peff)

Fig. 7 Dependence of stable power levels attained onpositive reactivity inserted.

From this study the safe operation specification isconfirmed , i.e. in order to avoid reactor shut-down byover-power condition it is necessary to limit positivereactivity inserted to value of less than 0.4

Furthermore, as already mentioned in Section 6, theTRIGA reactor has a stability behavior better than theDalat reactor because of TRJGA's higher negativeinstantaneous temperature coefficient. Indeed, it wasshown by Technical Foundation of TRIGA (9) thatthe TRIGA reactor can support inserted positivereactivity of 1 pe£p resulting in reactor period of 1 secand reaching only 120 kW stable power level afterhaving attained several hundred MW.

125

8 SURFACE TEMPERATURES OF FUELELEMENTS

Dalat reactor is a pool reactor cooled by naturalconvection. On the thermal hydraulic point of view,500 kW is a relatively high power for naturalconvection cooling. However, cooling was enhanced bythe addition of an extracting well during thereconstruction work (Fig 1).Water boiling at a fuelelement surface is harmful and can lead to a dangeroussituation. Consequently, the surface temperature of afuel element must be maintained well below thetemperature at which surfade boiling occurs. Therefore,careful measurement of fuel surface temperature canassist both nuclear safety and prospective powerupgrading studies.The device used to measure the fuel surfacetemperatures consists of an instnimented fuel element(TFE) together with an electronic measuring system.The IFE is of the same WR-M2 type . Nine Chromel -Coppel thermocouples (Tj -J-TQ) are incorporated inthe aluminum cladding of the hexagonal outer tubewith the different depths in the core ( Tt: 7 cm; Tj. 12cm; T3: 17 cm; T4: 30 cm; T5: 45 cm), two areincorporated in the aluminum cladding of the twocylindrical inner tubes ( T^: 9 cm, in the middle tubeand T7: 9 cm, in the innermost tube) and the remainingtwo others are in direct contact with water to measurethe core coolant inlet and outlet temperatures ( Tg: -4.5cm and Tg : 65.5 cm) .

Temperatures are measured at selected locations in thereactor core. At each of these locations, the normal fuelelement is replaced by the IFE and the thermocoupletemperatures monitored. At each location, 6 differentmeasurement sets could be obtained by revolving thehexagonal IFE around its axis by angle of 60°.

The axial fuel surface temperature distribution isdetermined for the outer hexagonal tube of the fuelelement at cells 9-6, 6-5 and 5-1. The measuredtemperatures Tj -f- T5 along the core height are shownin Fig 8 for various reactor power levels between 50kW and 500 kW. All curves have the same shape withthe maximum temperature position near the center ofthe core. The effect of shim rods is shown in Fig 8,where both the temperature distributions and maximumtemperatures vary slightly with shim rod depths.

The radial distribution of surface temperatures over thereactor core is obtained from measurements carried outat cells 5-1,6-3, 5-6 and 9-6. Fig 9 shows the T 4 data(i.e. maximum temperature at each fuel elementlocation) for a reactor power of 500 kW. It can be seenthat the highest T4 temperature is reached at one of thehottest positions in the core (cell 5-6) close to theneutron trap. The radial temperature distribution withinthe fuel element is obtained from data fromthermocouples T2, T5 and T7 which are located on thethree separate fueled sections of the IFE ( Fig 9 for cell

5-6). This figure indicates that the highest temperatureis on the outer surface near the neutron trap.

Temperature (°C)

100

90

80

70

60

50

40

30

P=500 kW

10 20 30 40 50 60depth (cm)

Fig. 8 Axial distribution of fuel surface temperatureat cell 9-6 for various values of reactor power,

o 4 shim rods at the same depth0 4 shim rods at difference depths

06.7 *c

neutron trnp

Fig. 9 Radial distribution of fuel surface temperaturea. Variation of temperature T4 with position at reactcpower of 500 kW. b. Temperatures T 2 , T$, T7 at eel5-4 at reactor power of 500 kW.

126

The data obtained at different locations in the coreindicate that cells 5-6 and 9-6 are at the hottestpositions, and the surface temperature at these positionsreaches a maximum of about 99°C on the hexagonalouter tube for the reactor operating at nominal power.This surface temperature coincides with the calculationvalue of 98.7°C, as shown by Ngo Phu Khang (10), andis much less than the value of 220°C of the TRIGAreaclor ( Technical Foundation of TRIGA, (9)).

Fig 10 shows the dependence on power level of fuelsurface temperature T4 and water temperature Tg atcells 5-6 and 6-3. It is clear that these temperaturesincrease proportionally with the reactor power at lowerpower levels and change more slowly at larger powerlevels.

Temperature (°C)

100

90

80

70

60

50

40

30

Cell 6-3

100 200 300 400 500P(kW)

Fig 10 Distribution of fuel surface temperature T4 andwater temperature Tg at cells 5-6 and 6-3 versusreactor power levels.

o Fuel surface temperature• Water temperature.

9 CONCLUSION

The main results of the investigation of the basicreactor physics characteristics of the Dalat nuclearresearch reactor are as follows :

a. The effective fraction of the delayed photoneutrons

induced by Be(y,n) reactions is 0.49 % peg- and theextraneous neutron source left after the reactor is shutdown gives the neutron flux of about (10^ 4- 10^)n/cm^/sec.

b. Lowest critical power levels of the reactor must bechosen greater than about 500 W due to the existenceof the extraneous neutron source.

c. The relative axial and radial distributions ofthethermal neutrons are measured. The geometricalparameters of the core are well agreed with thecalculation results.

d. The Dalat reaclor possesses an intrisicstabilitymuch less than the TRIGA reactor because theinstantaneous temperature coefficient of reactivity andthe safe positive reactivity inserted of the Dalat reactor( -0.02 10'4/°C and 0.4 Peff) are much less than thoseof the TRIGA reactor (-1.3 10"4/°Cand 1 pe f f).

e. The maximum fuel surface temperature atthehottest cell in the core can reach nearly 100°Cwhich is the safe limit for the Dalat reactor operating at500 kW nominal power. This surface temperature isless than that of the TRIGA reactor ( 220°C).

10 ACKNOWLEDGEMENTS

The authers are indebted to Prof. Pham Duy Hien, Dr.Tran Ha Anh, Dr. Vu Hai Long, Dr. Nguyen Tac Anhand Dr. Nguyen Phuoc Lan for usefull discussions. It isa pleasure to acknowledge the efforts of the reactor staffand the equipment assistance from the Nuclear PhysicsDepartment and Radiation Protection Department.

11 REFERENCES

1 Keepin S R, 1965, Physics of Nuclear Kinetics,Add Son-Wesley Publishing Company, Inc.

2 Krasin A K et al, 1958, Proceedings of 2ndGeneva Conference , Vol 12, p 571.

3 Loren C Schmid, 1971, Critical assemblies andreactor research, John Wiley & Sons, Inc.

4 Tong L S and Wei man J, 1970, ThermalAnalysis of PWR, ANS .

5 Murray R I, 1957, Nuclear Reactor Physics,Englewood Clifts, N.J. Prentice-Hall, Inc.

6 Technical design for reconstruction and expansionof the Dalat research reactor, 1979, Vol 3 (inRussian)

7 Nguyen Phuoc Lan, Ha Van Thong, Ngo QuangHuy, Do Quang Binh, 1992, Calculation of fuel burn-

127

up and fuel reloading for the Dalat nuclear research 9 Technical Foundation of TRIG A, 1958, GA-471.reactor. To be published.

10 NgoPhuKhang, 1987, Study of thermal hydraulic8 Directory of Nuclear ReactorsJ 959, Vol 2, IAEA, characteristics and possibility of power upgrading ofVienna the Dalat nuclear research reactor. Int. Rep., Dalat ,

Directory of Nuclear Reactors, 1960, Vol 3, IAEA, Vietnam.Vienna .

ABSTRACT The Dalat nuclear research reactor was reconstruced from theTRIGA Mark II reactor built in 1963with a nominal power of 250 kW The new reactor reached its planned nominal power of 500 kW for the first timein-February 1984. The undismounted components of the former reactor are the reactor tank, the concrete shielding,the graphite reflector, the beam tubes and the thermal column. The new components consist of the reactor core, thecooling system and the reactor control system. The new reactor core is loaded with the Russian-made WR-M2 fuelelements of uranium-aluminum aloy. An amount of beryllium material is placed at the center of the core, forming aneutron trap, and at the periphery of the core, forming a supplementary reflector.

The Dalat reactor has some characteristics distinct from the former TRIG A reactor. Investigation of itscharacteristics is carried out by the determination of the reactor physics parameters. The experimental results wereobtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor isshut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermalneutrons, the safe positive reactivity inserted into the reactor at a deep subcritical state, the reactivity temperaturecoefficient of water, the temperature on the surface of the fuel elements, etc.

KEYWORDS Triga type reactor, reactor physics, characteristics, neutron flux, thermal neutrons, distribution,reactivity, delayed neutrons, criticality, cooling system, temperature, fuel elements, control rods, reflector.

128

I l ium mm !•>» i««™ — — - —

VN9700006Proceedings of the 4th National Conference on Physics

Hanoi, Vietnam, 5 - 8 October 1993.

DELAYED PHOTONEUTRONSOF THE DALAT NUCLEAR RESEARCH REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Ho Chi Minh City, Vietnam.

Ha Van Thong, Vu Hai Long, Ngo Phu Khang, Nguyen Nhi Dien,Pham Van Lam, Huynh Dong Phuong, Luong Ba Vien, Le Vinh Vinh.

Nuclear Research Institute, Dalat, Vietnam.

ABSTRACT

Time spectrum of delayed neutrons of the Dalat nuclear research reactor ismeasured and analyzed. It corresponds to a shut-down neutron fluxes of about 10^-5- 10** n/cm^/sec after 100 hours continuous reactor operation at steady power levelof 500 kW. Data processing of experimental time neutron spectrum gives 16exponents, of which 10, resulting from photoneutrons due to (y,n) reactions onberyllium used inside the reactor core, are obtained by using successive exponentialstripping fitting method. For the Dalat reactor, the effective delayed photoneutronfraction relative to the total effective delayed neutron fraction is P e f f= 0.49 % f3efffor a beryllium weight relative to 11^35 fud of m3 e /my = 8.5. This result isacceptable in comparison to those obtained for other Be-U^-> media [1],[5].

1. INTRODUCTION.

The delayed neutrons in a reactor come from the two contributions : thedelayed fission neutrons and the delayed photoneutrons. The delayed photoneutronsare usually induced by (y,n) reactions due to the interaction of the gamma rays,emitted from the fission products, with heavy water or beryllium contained in thecore. The life-times of the delayed photoneutron precursors are generally greaterthan the life-times of the delayed fission neutrons. Therefore the dynamicproperty in region of the low frequency for the reactor moderated by heavy water orberyllium may have the higher inertial property in comparison to that of the reactorwithout photoneutrons [1].

The Dalai reactor belongs to reactors using 11^35 fuel, moderated and cooledby water. The core contains an amount of beryllium in a neutron trap and anadditional beryllium reflector. Therefore the determination of the photoneutroncontribution is useful in the study of the reactor kinetics and helps to estimate theintensity of the neutron source left after the reactor is shut down.

129

2. REACTOR CORE DESCRIPTION

The reactor core has a cylinder shape of 60 cm height and 43.6 cm radius and issurrounded by the old graphite reflector of 34.6 cm thickness. The core is loadedwith 89 fuel elements of WR-M2 36% enriched Ij235 type, 6 boron carbide controlrods (two safety rods and 4 shim rods), a stainless steel regulating rod, 3 irradiationchannels, a neutron trap and an additional beryllium reflector (Fig. 1). The neutron trapis placed at an area of 7 central fuel elements and is made of a beryllium block havinga water column of 65 mm diameter inside. The U^-^ weight in the core is 3.55 kg andthe beryllium weight is 30.2 kg, therefore the mass ratio is mgg/mTj = 8.5.

3. EXPERIMENTAL METHOD

Parameters of the delayed photoneutrons are determined by the reactor shut-down method [2]. It is proposed that the reactor is operating at NQ power and a 5knegative reactivity is inserted at T moment. So the power immediately decreases tothe Nj value according to the following expression:

m

where beff is the effective delayed neutron fraction, and after that it decreases withtime.

The time spectrum of the delayed neutrons may be analyzed by a sum ofexponential functions :

(2)

where a'j and V\ (i=l,...,6) are expansion constants for 6 groups of the delayedfission neutrons and a1; and A/j (j=l,...,n) are constants for the n photoneutrongroups. These constants are related to the partial effective fractions beff,i and beffjand decay constants \[ and X; (i=l,...,6 and j=l,...,n) by the following formulas[5]:

N(t)=

N0

6

ii=

*'i

-A.'e

i t+

nI

j=laJ

-Vjte

Peff?i(j) (j)— ( 1 - e ) (3)

Peff

orG)3eff + 5k

where

130

Cell 1-1

Fig. 1 Reactor core arrangementI. Fuel element; 2. Beryllium reflector 3. Neutrontrap ; 4. Wet channel ; 5. Pneumatic transfer Uibc ;6. Safety rod ; 7. Shim rod ; 8. Regulating rod ;9. Graphite reflector ; Cell 1-1 : The first cell of thecore, cell i-j denotes the cell at i*'1 row and j 1 ' 1 column.

10 20 50 t(min.)

10 20 30 40

o

Time spectrum of the delayed neutronsafter reactor shut dovm.

Experimental data.Fitting curve for 6 fission neutroncomponents and 10 photoneutron component:Calculation curve for 6 fission neutron

components.

131

Peff= 2 pcff,i + (5)

So using the experimental time spectrum and the least square fitting methodone may determine the parameters Peffj , Peffj » i anc* ^j- The w e ^ ^ow 1 1 [1]parameters Peffi a»d ^i can be used for checking the calculation method.Therefore in this work Peffj and \: are parameters to be determined.

4. EXPERIMENTAL RESULTS

The reactor shut-down experiment is carried out by a simultaneous insertion ofthe 4 control rods after 100 hours of continuos operation of the reactor at 500kW power. Then a negative reactivity of 5k = 1 ipeff is inserted.

The time spectrum of the delayed neutrons is measured by using the controlsystem, AKNP, of the reactor. The input information to the AKNP is given by the9 ionization chambers arranged in distances of about (60 + 77) cm from the corecenter. These 9 ionization chambers are composed of 3 groups responsible for 3ranges of the reactor power as follows :

Group

Source

Middle

Energy

Symbol

ID

PD

ED

Numberof Tonization

chambers

3

3

3

Power

10-8 o / o N ( )

10"3 % N 0 -

1 % N o

range

+ 10-2

*- 10%

4- 120

% N 0

N0

% N 0

Ionizationchamber

KNK-15

KNK-15

KNK-3

The measurement of the delayed neutron fluxes is carried out in a large rangeof 7 orders, so all the three types of ionization chambers are used. The timespectrum is registered into IN-90 MCA which has two inputs working in anuiltiscale regime. The first input collects information from a ID ionization chamber inthe first 30 sec up to 1% NQ power level and after that information from a PDionization chamber. The second input collects information from a PD ionizationchamber started from 0.5 sec after the reactor shut-down and stops its collection atthe power of 10'^ % NQ. This procedure allows to combine information fromthree ionization chambers and gives a continuous time spectrum in 64 hours duration(Fig. 2).

132

Table 1. Experimental data on the delayed neutrons measured by differentmethods (relative to the power level before reactor shut down).

Time afterreactor shut-

down

35.7 min.42.4 min.49.75 min.57.25 min.

1.09 h1.83 h10.03 h13.27 h20.49 h28.95 h30.5 h40.48 h40.58 h49.56 h50.2 h59.98 h60.93 h

Data obtained by various measurement methods

From ionizationchambers

(3.9 ± 0.2) 10"5(3.6 ±0.2) 10-5(3.3 ± 0.2) 10*5(3.1 ±0.2) 10*5(2.9 ±0.2) 10-5(2.2 ± 0.2) 10"5(4.5 ± 0.3) 10-6(3.1 ± 0.2) 10-6(1.8 ± 0.1) 10"6(1.1 ± 0.1) 10'6(1.00±0.08)10-6

(5.9 ±0.5) 10-7(6.0 ±0.5) 10'7(4.8 ± 0.4) 10'7(4.7 ± 0.4) 10-7(4.1 ± 0.4) 10-7(4.0 ± 0.4) 10-7

Activation ofindium foils

(3.5 ± 0.2) 10-5(3.4 ± 0.2) 10"5(3.0 ± 0.2) 10-5

(2.9 ±0.1) 10-5{2.7± 0.1) 10"5

(4.4 ± 0.2) 10"6

(1.8 ± 0.2) 10"6

(0.95±0.05)10'6(6.2±0.4) 10-7

(5.0 ±0.3) 10"7

(4.3 ±0.2) 10-7

activation ofcopper foils

(2.3 ± 0.2) 10-5

(3.2 ± 0.3) 10"6

(1.2 ± 0.1) 10"6

(6.1 ± 0.4) 10"7

(4.6 ±0.3) 10-7

(4.2 ± 0.3) 10-7

The ionization chambers work in a gamma compensation regime forregistering neutrons in the field of high density gamma rays, mainly in the very lowpower levels after reactor shut-down. In order to check the gamma compensationlevel of ionization chambers, it is useful to measure the delayed neutron spectra byindium and copper foils activation on the base of In^^(n,y)In^^m andCu63(n,y)Cu64 reactions respectively. I n ' ^ m has two important picks of 1097.2keV and 1293.5 keV with the branch coefficients of 57.3 % and 84.4 %respectively. Cu^4 has the picks of 511 keV and 1345 keV with the branch coefficientsof 37 % and 0.5 % respectively. Indium foils are irradiated at dry 7-1 channel byusing a pneumatic transfer system. Copper foils are irradiated at the neutrontrap. Equipment for measuring gamma spectra is GeHP detector of 70 cm-* volume inconnection to CANBERRA MCA. Table 1 presents the results obtained by using theionization chambers and by activating the indium and copper foils.

From Table 1 it is seen that the experimental data from the 3 differentmeasurements are well agreed with each other within error ranges, that means agood compensation of ionization chambers.

5. ANALYSIS OF EXPERIMENTAL DATA

From Fig. 2 it is obvious that the delayed neutron spectrum decreases rapidlyin the first 10 minutes and after that slower decreases and remains about 10"^ -s-_ ^

133

of nominal power during 60 hours. Calculation curve of delayed U^35 fissionneutrons (Fig.2) shows that at 10 minutes after reactor shut-down the counts ofdelayed fission neutrons may be neglected in comparison to those ofphotoneutrons. In consequence, the neutron source left after reactor shut-downis practically composed of the only delayed photoneutrons and has neutron fluxdensities of about (10^ -*- 10**) n/cm^/sec.

Parameters in the expression (2) for the delayed neutron groups are determinedfrom the experimental time spectrum by using the least square fitting method [3]. Inreference [1] the time spectrum of the delayed neutrons composed of 15 exponents,of which 6 exponents correspond to the delayed fission neutrons and 9 - the delayedphotoneutrons. The life times of the precursors of the delayed neutrons lie in therange from 0.2 sec to 12.8 days, therefore it is difficult to fit all of the parameters atthe same time. This difficulty is overcame by using a successive exponential strippingmethod, in which the time spectrum of the delayed neutrons is divided into thesections having the equal approximation values of precursor's life times. Dataprocessing is carried out section by section started from the right end of the spectrumto the left.

Table 2 shows the parameters of 6 groups of the delayed fission neutrons andtable 3 presents 10 groups of the delayed photoneutrons obtained by the abovementioned method of the spectrum analysis. From table 2 it is obvious that if Peff=

(5 = 0.66 % ( = 0.64 %± 15.17 10-5 ) then partial fractions |3eff j (i=l,2,...,6) of thiswork well agree with those of G. R. Keepin's data [1] within error ranges. It meansthat the used procedure of analysis is available. The spectrum analysis allows to obtain10 delayed photoneutron groups, 8 of which have the life times coincided with those ofKeepin's data (Table 3). The 2 groups of 6.8 hours and 2.8 hours life times of thiswork (see reference [4]) correspond to the group of 3.11 hours life time in reference

Table 2. Parameters of 6 delayed fission neutron groups.

No

123456

E(KeV)

250560430620420

-

Total

T l /2(sec)

55.722.76.22.30.60.2

Peff,i x 103

(Peff)

33 ± 3219 ± 18198 ± 13385 ± 20117± 842 ± 2

(99.5 ± 3)

% Peff

Peff,i * 105

22 ± 2143 ± 12130 + 9

252 ± 1377 ± 528 ± 1

(0.65 ± 0.02)%

PJXIO5

211401252537427

0.64 %

134

Table 3. Parameters of 10 delayed photoneutron groups.

N0

12345678910

Precursors

Ba-140Te-132

Te-131 and others1-135

Kr-88, La-142, Kr-87Te-134, 1-134, Br-84

Xe-138, Rb-89, Mo-101Sb-133, Kr-89,Rb-90

1-136, Br-87Kr-90, Se-87

Tl/2j

12.8 d77.7 h12.1 h6.8 h2.8 h

43.2 min15.5 tnin3.2 min1.3 min

0.51 min

PBeeffj 10*(PefF)

1.54 ± 0.120.62 ± 0.041.95 + 0.139.81+ 0.7618.6 ± 1.046.5 ± 2.020.6 ± 1.2

112± 9146 ± 11136 ± 11

T l /2Keepin [1]

12.8 d77.7 h12.1 h3.11 h3.11 h

43.2 min15.5 min3.2 min1.3 min

0.51 min

Pj * 105Keepin [1]

0.0570.0380.2603.203.200.363.681.853.662.07

Total effective delayed neutron fraction beff consists of two parts: effeffective delayed fission neutron fraction and P^eeff, effective delayed photoneutronfraction. Experimental data of delayed neutron spectra have been time analyzed [1]into 15 groups, of which 6 groups are i P - ^ fission and 9 others from (y,n)reaction on beryllium. It resulted that in media composed only of beryllium,delayed photoneutron fraction is 15.175 10" , i.e. with known delayed neutronfraction from Ij235 fission pU = 0.0064, one has

= 2.31 % ; pU = 97,69 %

In an experiment on rapid shutdown realized on a critical assembly withcylindrical core composed of enriched uranium (10% U^35) and berylliumreflector, (mge/inij - Mil -*- 3112 ) [2], one obtained delayed photoneutronfraction = 14.808 10"5, so

P B e = 2. 26 %; pU = 99.74 %

These data show that the effective fraction of delayedpliotoneutrons increases with mgg/iriTj ratio as illustrated in Table 4.

1 able 4. Dependence of P^eflp'Peff o n mBe^mU r a t '°

mBe/mTj

8.51777 -=- 3112

Pure beryllium

PBeeff'Peff

0.49 %2.26 %2.31 %

Reference

This work[5][1]

135

6. CONCLUSION

The main result of the investigation of the delayed photoneutrons in the Dalatreactor is as follows :

1. The time spectrum of the delayed neutrons is determined in 64 hours durationafter the reactor is shut down following 100 hours of its continuos operation.

2. For about 10 minutes after reactor shut-down, the delayed neutrons arecomposed of the only delayed photoneutrons induced by the interaction of the gammarays, emitted from the fission products, with beryllium contained in the core. The fluxdensities of the delayed photoneutron source are about (10^ ~ 10**) n/cm^/sec.

3. The time spectrum of the delayed photoneutrons is analyzed into 16 componentsincluding 6 components of the delayed fission neutrons and 10 components of thedelayed photoneutrons.

4. The total effective fraction of the delayed photoneutrons is (3®eeff = 0.49 %

Peff Th's result is acceptable in comparison to those obtained for other Be - Ij235media.

7. ACKNOWLEDGMENTS

The authors are indebted to Prof Pham Duy Hien, Dr. Tran Ha Anh, Dr. NguyenTac Anh, Dr. Pham Ngoc Chuong, Dr. Tran Khanh Mai for useful discussions.- It is apleasure to acknowledge the efforts of the reactor staff, and the equipment assistancefrom the Nuclear Physics Department and Radiation Protection Department.

8. REFERENCES

1. G.R. Keepin. Physics of nuclear kinetics.Addison-Wesley Publishing Company, Inc., 1965.

2. F.G. La Violette. Naval reactor physics handbook. Vol. Ill, Chapter 25.USAECJ959.

3. Z.Szatmary. Data evaluation problems in reactor physics. Theory of program RFIT.Report KFKI - 1977 - 43, Budapest.

4. D.L. Hetrick and W.W. Bwown. Trans. Amer. Soc. 3, N 2, 435, 1960.5. A.K. Krasin et al. Proceeding of 2nd Geneva Conference. Vol. 12, 571, 1958.

136

•I1BIIB1 • • • iput • • • « • • • • • « • • • • — •• ——- ••

VN9700007Proceedings of the 4th National Conference on Physics

Hanoi, Vietnam, 5 - 8 October 1993.

DETERMINATION OF THE LOWEST CRITICAL POWER LEVELSOF THE DALAT NUCLEAR RESEARCH REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Ho Chi Minn City, Vietnam.

Ha Van Thong, Vu Hai Long, Do Quang Binh, Huynh Ton Nghiem,Nguyen Minn Tuan, Luong Ba Vien, Le Vinh Vinh.

Nuclear Research Institute, Dalat, Vietnam.

ABSTRACT

This paper presents the experimental methods for determining critical statesof the Dalat nuclear research reactor containing an extraneous neutron sourceinduced by gamma ray reactions on beryllium in the reactor. The lowest criticalpower levels are measured at various moments after the reactor is shut downfollowing 100 hours of its continuous operation. The power levels vary from (0.5 -1.2)xKrPn, i.e. (25-60) W to (1.1 - 1.6)x 10"5Pn, i.e. (5.5 - 8) W at correspondingtimes of 4 days to 13 days after the reactor is shut down. However the criticalpower must be chosen greater than 500 W to sustain the steady criticality ofthe reactor for a long time.

1. INTRODUCTION

The Dalat nuclear research reactor was reconstructed from the former TRIGAMARK II reactor. In the new core is placed about 30 kg of beryllium material thatinduces an extraneous neutron source via Be(gamma,n) reaction [1]. Thepresence of this extraneous neutron source increases the lowest critical powerlevels of the reactor. Determination of the lowest critical power levels isinteresting from the view-point of safety in the reactor start-up process andfrom the requirement of finding possible low critical powers for the reactor physicsexperiments.

In this paper are presented the method for determining the reactor criticalityand the lowest critical power levels of the Dalat nuclear research reactor.

2. EXPERIMENTAL METHOD

It is assumed that the reactor should be kept at a subcritical state with amultiplication factor ko <1. Then the asymptotic neutron density No is related to

137

the neutron density of the extraneous neutron source Ns by the following formula :

No = NsA5ko (1)where c5kn = 1 - ko (2)

c5o is the excess multiplication factor of the subcritical reactor.

After the insertion of a positive reactivity p, the neutron density increases andreaches a new asymptotic value. The il and (i+1)11 insertions of a positive reactivitygive the following asymptotic subcritical neutron densities :

Ni = Ns/<5ki (3)N|+i = Ns/c5ki+l (4)

where 6 kj = 1 - (ko + \p) -6 ko - \p (5)c3ki + i =c5kn-(i + 1> (6)

The (i-t-2)1 insertion of positive reactivity gives the new state of the reactorwith the new density Nj+2. This new state is subcritical if Nj+i < 2Nj, critical ifNi+i = 2Nj, and supercritical if Nj+i> 2Nj [2]. In the case of Nj+i > 2Nj, theneutron density Nj+2 at first slowly increases in subcritical state and after passingthe criticality. It has an exponential form in the supercritical state :

Log N(t) = const + t/T (7)

where T is the reactor period. So in semilog plot, the beginning moment of thelinear function (7) is the moment of reactor criticality.

3. SUBCRITICAL STATES OF REACTOR

The experiments are carried out in March, 1992 at 10 days after the reactorwas shut down following 100 hours of its continuous operation . The twosafety rods are withdrawn completely, the four shim rods are placed at 295 mmin depth and the regulating rod is at 350 mm. The reactor power is 6.7x10" Pn,where Pn is the 500 kW nominal power. The reactor power is increased by shootinga cadmium sample from the reactor core using a pneumatic transfer system. Itis equivalent to the insertion of a positive reactivity of p — 4.67%/?eff, where /?effis the total effective delayed neutron fraction.

Fig. 1 shows the neutron density response to the positive step changes inreactivity. The density curves in subcritical states increase and reach the asymptoticvalues. They are well described by the subcritical reactor kinetics theory [3] withthe positive step changes in reactivity of p = 4.67%

Table 1 presents the asymptotic neutron densities Nj, the ratios Nj/Nj-i andthe reactor powers Pj for 6 reactivity insertions. From Table 1, it is obvious thatN6/N5 > 2, so in the 7 l insertion of the 4.67% 0eff positive reactivity, the reactorwill reach the supercritical state.

138

21OO

2000

13OO -

12OO

6k»3. 04<x

+ •

2O 4O 6O SO 1OO

Tim© (xiOoec)

12O 16O

6OOO

55OO

50O0

35OO

3OOO

25OO

2.OOO

<5k-1.04a

O 2O 4O 6O 8O 1OO 12O 1-4O 16O 18O 2OO

Tlrr>» (x1O8ec)

Fig. 1. Subcritical neutron density response to the positivestep changes in reactivity Ca » 4.67X/?©fO

+ : Experimental data CMarch 16th, 199SD,- - - : Calculation with 6 delayed neutron groups,

: Calculation with 16 delayed neutron groups.

139

The subcriticnl excess multiplication factor c3ko and the neutron density ofextraneous source Ns are determined by the least square fit method for thefunction :

Ni =- (8)- i

i= 1,2,3,4,5,6.

The obtained results are as follows :

<5ko = (6.041 ±0.018 )P

Ps = (1.93 ±0.01)xl0"8Pn

where Ps is the power corresponding to the extraneous neutron source

Table 1. Neutron densities and reactor powers in subcritical states.

i

0

1

2

3

4

5

6

Nj (cps)

945 ± 5

1120 ± 10

1400 ± 10

1840 ± 10

. 2792 ± 10

5426 ± 20

Nonasymplolic value

Ni/Ni + 1

1.185 ±0.002

1.250 ±0.002

1.314 ±0.002

1.517 ±0.(K)2

1.193 ±0.003

> > 2

P^IO-VJ

0.850 ±0.001

1.01 ±0.01

1.26 ±0.01

1.66 ±0.01

2.52 ±0.01

4.88 ± 0.02

4. DETERMINATION OF THE LOWEST CRITICAL POWER LEVELS

The criticality of the reactor is determined by withdrawing continuously theregulating rod at a speed of 2.8 mm/sec from the depth of 400 mm to 200 mm whilethe four shim rods are located at the depth of 284 mm. The criticality is obtainedin semilog LogN(t) function (Fig. 2) by determining the starting point of the linearpart of the curve.

The lowest critical powers depend on the beginning positions of the regulatingrod. The obtained values are in the range 1.6x10"' Pn to 4.2x10"' Pn for thebeginning positions of the regulating rod from 400 mm to 600 mm.

The range of the lowest critical powers depends on the measurementmoment after the reactor is shut down. Fig. 3 shows the ranges of the lowest criticalpower levels as a function of time. They have the values from (0.5 - 1.2)xlO" Pn,i.e. (25 - 60)W to (1.1 - 1.6)xlO"5 Pn, i.e. (5.5 - 8)W at corresponding times of 4 daysto 13 days after the reactor is shut down.

In the just critical situation, when the multiplication factor of the reactork = 1, the neutron density or power level will rise at a linear rate [3] :

140

-)O3

1O2

-101

1 0 0 >OO 3OO

Time <sec)

4 O O 5 GO

ig Neutron dens ty response to the passing of he reactorfrom the subcritical state to the supercritical state.

-a

a.

O"

1O'

1O

(Watt)

500

100

10

3 4 5 6 7 8 9 1O 1 1 12 13 1-4 15 16Time after reactor shut down (day)

Fig 3. Lowest critical power levels as function of time afterthe reactor is shut down.

1 : Experiment,: Calculated value of Pc so that the first term on

the right hand of the formula (9) equals 300% (a), 100%<b) , 10% (c) and 5% id) of the second one during

141

5OO r

1OO 2.OO 3OO 4OO 5OO ©OO 7OO SOO <=>OO 1OOO 1 1OO

Fig 4. Linear increase of the reactor power at the criticalstate due to the existence of an extraneous neutronsource.

o :

AR = 310 mmAR = 320 mmAR = 330 mmCalculation

142

P(t) = 1 + Pc (9)

vvliere 1 : prompt neutron lifetime, /?m, m : effective delayed neutron fraction anddecay constant of the m1 * neutron group, Pc : critical power of reactor withoutextraneous source.

At the moment of 10 days after the reactor is shut down, the formula (9) can beexpressed as follows :

P(t) = 5xl0"'SPn t + Pc (10)

In Fig.4 is illustrated the expression (10) together with the experimental data,where Pc = 1.6x10"* Pn is considered as 100% at the starting moment t=0. FromFig.4 it is seen that during 10'sec the first term on the right hand of the formula(10) is greater than the second one by 300%. It means that the reactor power isnot just constant but rapidly increases. In order to decrease the power variationsmaller than 10%, it is necessary to choose the critical power level of 5x10" Pn.

In Fig. 3, the critical powers are plotted versus the time T after the reactor isshut down so that the first term on the right hand of the formula (9) equals 300%(curve a), 100% (curve b), 10% (curve c) and 5% (curve d) respectively of thesecond one during the 10" sec. From Fig. 3 it is obvious that the lowest criticalpower levels measured vary in the range of (100 - 300)% during the first 10\sec.

The power variation of (100 - 300)% are too large for practical use of reactor.They must be reduced to the values of about (5 - 10)% to sustain the steadycriticality of the reactor. Therefore, the critical powers must be chosen greater than10"3Pnor500 W.

ACKNOWLEDGEMENTS

The authors are indebted to Prof. Phnm Duy Hien and Dr. Tran Ha Anh forinterest in the work and discussion. It is a pleasure to acknowledge the reactor stafffor help in experiment and the Nuclear Physics Department for the use ofpneumatic transfer system.

REFERENCES

1. Nguyen Nhi Dien, et al. 'Delayed photoneutron components of the Dalat nuclearresearch reactor', Preprint DNRI, Dalat, V 034-89, 1989.2. V. I. Vladimirnv.'Practice exercises on exploitation of nuclear reactnrs\Moscow,Energoizdat, 19X1 (in Russian).3. Loren C. Schmid.'Critical Assemblies and Reactor Research', John Wiley andSons, Inc., 1971.

143

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993. VN9700008

THE NEUTRON FIELD PERTURBATION EFFECTIN THE DALAT REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Ho Chi Minh City, Vietnam.

Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh.

Nuclear Research Institute, Dalat, Vietnam.

ABSTRACT

The perturbation effect of the thermal neutron field of the Dalat nuclearresearch reactor is investigated when a fuel element is replaced by a water column or aplexiglass rod. In consequence, it is possible to replace the measurement of therelative distribution of the thermal neutron field on the surface of fuel element by that inthe water column or in the plexiglass rod.

1. INTRODUCTION

The relative distribution of the thermal neutron field in the reactor core is oftenmeasured by the method of neutron activation of foils placed on the surface of fuelelements . This method of the measurement was carried out at power levels of a fewwatts in the physical start-up stage of the Dalat reactor but can not be applied in thecourse of operation of the reactor because the existence of the delayed photoneutronsource [1] keeps the minimum critical power level at about 500 watts [2]. Therefore,finding another condition for measuring the distribution of the thermal neutron field inthe core of Dalat reactor is necessary.

According to [3], in a heterogeneous reactor a replacement of the fuel element byanother material causes a perturbation which creates a change of the distribution of thethermal neutron field in the core in general, and at the location of the fuel element, inparticular. However, the relative change of both the thermal neutron density NQ onthe surface of fuel element and the thermal neutron density N in the replaced material

A = ( N 0 - N ) / N 0 (1)

is constant, that does not depend on the position of the fuel element in the core butdepends only on the type of replaced material. This conclusion will be right if the size ofthe fuel element is very small compared with the size of the core and if theperturbation causes a small change of the multiplication factor k of the reactor.

Constancy of A reflects a similarity of the relative distribution curves of thethermal neutron field on the surface of fuel element and in the replaced material. In

145

Cell 1-1

Fig. J Reactor core arrangement1. Fuel element ; 2. Beryllium reflector 3. Neutrontrap; 4. Wet channel ; 5. Pneumatic transfer tube ;6. Safety rod ; 7. Shim rod ; 8. Regulating rod ;9. Graphite reflector ; Cell 1-1 : The first cell of thecore, cell i-j denotes the cell at i"1 row and j t n column.

z.s0.9 QJ 0.3

\

Al U-Al

Cross section of the fuel element( Unit in mm )

Fig. 3 Cross section of the plexiglass re

146

consequence, it is possible to replace the measurement of the relative distribution of thethermal neutron field on the fuel element surface by that in the replaced material.

For the Dalat reactor, the size of the fuel element, about 3.2 cm , can not beconsidered very small compared with the size of the fuel area of the core with 18.9 cmouter diameter and about 9 cm inner diameter (the central neutron trap). Therefore, aexperimental check of the neutron field perturbation and the constancy of the coefficientA is necessary for using the above mentioned method of measuring the relativedistribution of thermal neutron field.

2. PERTURBATION EFFECT WHEN THE WATER COLUMN ISREPLACED BY THE PLEXIGLASS ROD.

2.1 Experimental method.

The Dalat reactor was reconstructed from the former reactor TRIGA Mark II , inwhich the new core was installed inside the old graphite reflector . The core (Fig. 1)contains 89 fuel elements, 7 control rods (2 safety rods AZj, AZ2, 4 shim rods KCj -*-KC4 and 1 automatic regulating rod AR), 1 neutron trap, 2 pneumatic transferchannels (at cells 7-1 and 13-2) and 1 "wet" irradiation channel (at cell 1-4). A berylliumreflector layer was installed between the core and the graphite reflector. Each fuelelement consists of three coaxial tubes, in which the outside tube has a hexagonal crosssection and the other tubes have circular cross sections (Fig. 2). The fuel element has anoverall height of 86.5 cm , in which the fuel part is 60 cm high with 36 % uraniumenrichment.

The coefficient A is determined for the core configuration of 85 fuel elements aftertaking away the fuel elements at cells 4-1, 7-3, 6-8, and 9-5. Hence, together with"wet" irradiation channel 1-4 we have five water columns. Determination of the axialdistributions of the thermal neutron field at the five water columns in the core iscarried out by neutron activation of Cu foils at power level of 2.5 kW for 15 minutes.The Cu foils are equally spaced 5 cm apart on one side of a thin plexiglass plate situatedin the center of the water column. After that, at each cell the water column is replaced by asolid plexiglass rod of 80 cm high with the same outside size as the fuel element (Fig.3).Determination of the axial distribution of the thermal neutron field along the plexiglassrod is also carried out by neutron activation of Cu foils placed along the center lineof the rod. Activity of the Cu foils is measured by beta counters. Reactor powers fordifferent measurements are calibrated by neutron activation of Cu foils in the rotaryspecimen rack and the total counts of the three ionization chambers of the reactorcontrol system.

The axial distributions of the thermal neutron field in the water column and in theplexiglass rod are described by the following expression [4]:

O(z) = [ C + D(z-30) ]cosB(z-30) (2)

147

K ('cm )

Fig. 4. Axial distributions of the thermal neutron field at the cells I-4, 7-3 and 6-8.o : In water column+ : In plexiglass rod

148

where C, B, D are constants, z (cm) presents the reactor height counted from the top ofthe reactor.

When the reactor operates at power level of 2.5 kW, the depth of the shim rodsKC] •*• KC4 remains unchanged, while the depth of the regulating rod AR is at 34.5 cmfor the water column and changes from 36 cm to 38 cm for the plexiglass rod. So, theshapes of the thermal neutron distribution curves are different. The coefficient A iscalculated at the depth of z = 30 cm so that it does not depend on z. Then

A = ( N o - N ) / N o = ( C o - C ) / C o (3)

where NQ, CQ and N, C are the constants for the water column and the plexiglass rodrespectively.

2.2 Similarity of the thermal neutron distribution curves for the water column andthe plexiglass rod.

The similarity is verified at cell 6-4 in the core configuration of 88 fuelelements. The depths of the shim rods KCj-r- KC4 and the regulating rod AR have thesame values in the two measurements for the water column and the plexiglass rod. Thetwo thermal neutron distribution curves are described by the expression (2). Theconstants C, B, D for the plexiglass rod and CQ, BQ, DQ for the water column aredetermined with ±5% error and equal

C = 1.485 D = 1.934 10"2 B - 3 . 9 5 10"2

C o = 0.910 D Q = 1.198 lO-2 B 0 = 4.02 10-2

From the above data it is obvious that the coefficients B and BQ are coincident inerror range, while the ratio C/CQ = 1.63 ± 0.11 equals ratio D/Drj = 1.61 ± 0.11 in errorrange. So the two thermal neutron distribution curves for the water column and theplexiglass rod are similar.

2.3 Coefficient A for cells 1-4, 4-1, 7-3, 6-8 and 9-5 when the water column isreplaced by the plexiglass rod.

Figure 4 presents the experimental results and calculation curves according to theexpression (2) for the axial distribution of the thermal neutron field at cells 1-4, 7-3 and6-8. From this figure we find that the thermal neutron field in the plexiglass rod islower than in the water column. On the other hand, the shapes of the pairs of thedistribution curves arc not coincident because the depths of the regulating rod ARare different for different measurements. The coefficient A is calculated at the depth of7 = 30 cm according to the expression (3) . Table 1 presents the calculated results ofthe constants C, CQ together with the coefficient A.

149

Table 1. Measured values of the constants C, Co and A (in relative unit).

Cell1-44-16-87-39-5

Co (Water)1.142 ± 0.0111.116 ± 0.0221.715 ± 0.0161.247 ± 0.0151.490 ±0 .016

C (Plexiglass)1.038 ±0.0111.021 ± 0.0041.561 ± 0.0141.133 ± 0.0071.362 ±0.015

A0.091 ± 0.0130.085 ± 0.0180.090 ± 0.0120.091 ± 0.0120.086 ± 0.011

From Table 1, it is seen that the coefficients A for the five investigated cells havethe same values in error range. In the five cells, two cells 1-4 and 4-1 lie next to theberyllium reflector, two cells 6-8 and 9-5 lie next to the beryllium block of the neutrontrap and cell 7-3 lies in the middle of the core. Therefore, the coefficient A is constant,not depending on the position of the cells in the core. For the Dalat reactor, thisconclusion is of great significance because the size of a fuel cell cannot be consideredvery small compared with the size of the core.

2.4 Neutron field perturbation effect whenplexiglass rod.

the water column is replaced by the

According to [3], the coefficient A is constant only when the replacement ofthe water column by the plexiglass rod creates a perturbation. To perform this, inmeasurements we must keep power levels unchanged, that is , to ensure a smallchange of the multiplication factor k of the reactor. In this work, the perturbation isverified by using the two following methods . Because the two methods are independent,it is possible to verify the perturbation effect by one of them.

1. The first method: Replacement of the water column by the plexiglass rodintroduces a small positive reactivity p into the reactor and according to the perturbationtheory [5], this reactivity is proportional to the neutron density square at the perturbedcenter. This proportion has the form

HF = J CD2(Z) dz

0(4)

Table 2 presents the reactivity p, measured by using the regulating rod AR , thevalue F and the ratio F/p for the cells 4-1, 6-8 and 7-3. From Table 2 it is seen that theratios F/p have the same values in error range, that is, the proportion in the expression(4) is satisfactory.

150

Table 2. The experimental data for p, F and F/p .

Cell

P({3ftflr)±io%F ± . 5%F/p ± 12%

4-11.9 10"2

23.612.4 102

6-84.5 10"2

56.012.4 102

7-32.5 10"2

29.111.6 102

2) The second method: In [3],it is found that perturbation might cause a strongchange of the neutron field at the position of the replaced material, but no significantchange in the neutron field at the other places in the core. This is performed in this work.Indeed, when the water column is replaced by the plexiglass rod at the five investigatedcells, the thermal neutron density changes fairly greatly, about 10%, see Table 1, butthe thermal neutron densities at the neutron trap and the cell 1-4 almost do not changewith a such replacement at the cells 4-1, 7-3, 9-3, Table 3.

Table 3. Neutron densities at the neutron trap and the cell 1-4.

Experiment

Water column at cells 4-1,7-3 and 9-5Plexiglass rod at cell 4-1Plexiglass rod at cell 7-3Plexiglass rod at cell 9-5

Neutron density (relative unit)At the neutron trap

2.31 ± 0.092.26 ± 0.092.11 ± 0 . 0 82.43 ± 0.10

At the cell 1-4

1.35 ± 0.061.35 ± 0 . 0 61.30 ± 0.051.36 ± 0.05

3.' PERTURBATION WHENTHE PLEXIGLASS ROD.

THE FUEL ELEMENT IS REPLACED BY

The measurement of the neutron field distribution on the surface of the fuelelements by neutron activation of foils is impossible because of the high critical powerlevels of the reactor as indicated in [2]. Therefore the constancy of the coefficient A isunable to be directly verified when the fuel element is replaced by other material.However, according to [3], this constancy is right if the replacement of the fuel elementby another material creates perturbation of the neutron field . Then the perturbationcan be verified by the second method presented in 2.4. In the experiment, the thermalneutron densities at the neutron trap, the cell 1-4 and some positions of the rotaryspecimen rack are determined when replacing the fuel element by the plexiglass rod atcells 4-10, 5-10, 6-10, 6-9 and 6-8 (Table 4). From Table 4 it is seen that the thermalneutron densities at the considered positions are coincident in error range.

151

Table 4. The thermal neutron densities at the neutron trap, the cell 1-4 and somepositions in the rotary specimen rack (R. S. R) when replacing the fuel element by theplexiglass rod .

Positions ofplexiglass rod

4-105-106-106-96-8

-

Number offuel elements

888888888889

Thermal neutron density (relative unit)Neutron trap

3.72 ± 0.043.71 ± 0.093.72 ± 0.103.68 ± 0.083.66 ± 0.08

-

Cell 1 - 41.19 ± 0.041.20 ± 0.041.23 ± 0.041.21 ± 0.041.19 ±0 .041.24 ± 0.04

R. S. R.1.05 ± 0.031.08 ± 0.041.08 ± 0.041.09 ± 0.041.07 ± 0.041.07 ± 0.04

4. CONCLUSION.

In the paper, the perturbation of the thermal neutron field is investigated when thefuel element is replaced by the water column or the plexiglass rod . The results are asfollows:

1- The axial distributions of the thermal neutron field both in the watercolumn and in the plexiglass rod are determined at the cells 1-4, 4-1, 7-3, 6-8, and 9-5.A similarity of the distribution curves of the thermal neutron field between the watercolumn and the plexiglass rod is obtained. The coefficients A for the five consideredcells are coincident in error range.

2- The replacement of the water column by the plexiglass rod or the replacementof the fuel element by the plexiglass rod cause the perturbations of the neutron field in thecore. In consequence the measurement of the relative distribution of the thermal neutronfield in the core may be carried out by the neutron activation of foils in the water columnor plexiglass rod instead of on the surface of the fuel element.

5. ACKNOWLEDGMENTS.

The authors are grateful to Professor Pham Duy Hien, Dr. Tran Ha Anh, Dr.Nguyen Tac Anh, Professor Cao Chi, Dr. Pham Ngoc Chuong, Dr. Nguyen PhuocLan, Dr. Tran Khanh Mai, Dr. Nguyen Tien Nguyen for their helpful comments.Theauthors also thank the reactor operation shifts for helping to perform the experimentsand thank Nuclear Physics Department , Radiation Safety Department for makingconditions comfortable to use experimental equipment.

152

6. REFERENCES

1 Nguyen Nhi Dien, Ngo Quang Huy et al,Delayed photoneutrons of the Dalat nuclear research reactor,Preprint of the Nuclear Research Institute, Dalat, Vietnam, V 034-89, 1989.

2 Ngo Quang Huy, Ha Van Thong et al,Determination of the lowest power levels of the Dalat nuclear research reactor,National Conference on Physics, Hanoi, Vietnam, October 1993.

3 A. D. Galanin, Theory of heterogeneous reactors,Moscow, Atomizdat, 1971 (in Russian ).

4. L. S. Tong and J. Weiman, Thermal analysis of Pressurized Water Reactor.American Nuclear Society, 1970, chapter 1.

5. R. I. Murray, Nuclear Reactor Physics,Englewood Cliff, N. J. Prentice-Hall, Inc. 1957.

153

Hi s-AC-'-t

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993. VN9700009

EFFECTIVE HEIGHT OF THE CORE

OF THE DALAT NUCLEAR RESEARCH REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Ho Chi Minh City, Vietnam.

Ha Van Thong, Vu Hai Long, Nguyen Due Binh,Nguyen Minh Tuan, Luong Ba Vien, Le Vinh VinhNuclear Research Institute, Dalat, Vietnam.

David Perez Martin, Fernando Garcia YipHigh Institute of Nuclear Sciences and Technology, Cuba.

SUMMARY

Measurements of thermal neutron relative distributions inaxial direction at different positions in the reactor core and forvarious control rod configurations have been carried out, andaxial buckling and effective height of the core deduced.

1. INTRODUCTION

The Dalat nuclear research reactor, renovated from the TRIGA MARK IIreator, has a modified core structure designed to run with Russian fuel element ofVVR-M2 type. The net height of fuel element containing uranium is 60cm, while theeffective height depends on material structure of the upper and lower parts of thecore. The measurement of the core effective height is significant from experimentalas well as theoretical viewpoints for the reason that the buckling of the reactor, animportant parameter for reactor computation, can also be deduced. For thisdetermination, the relative distribution of thermal neutron versus vertical distance isto be measured. However, the form of this distribution strongly depends on controlrod configuration inside the core, therefore a precise measurement of effective heightrequires a set of such distributions. The experimental data set, in turn, cancontribute significaltly to the optimization of sample irradiation, as well as permitthe comparision with theoretical neutron flux calculation.

In the following are presented experimental results on axial neutron distributionstogether with deduced values of axial buckling and effective height of the core.

2. CORE CONFIGURATION

The reactor core has cylindrical form with 43.6cm diameter surrounded by a34.6cm thick graphite reflector (Fig. 1). The present loading is composed of 89 fuelelements. The uranium-aluminum alloy fuel element with 86.5cm total length is made

155

Fig.l- Horizontal section of reactor core

1- neutron trap, 2- beam tube, 3- graphite reflector£) fuel element,^} (££& Be shirn element and reflector-Irradiation channels! (Q) 1-4, (jjij) 7~1» ®) 13-2Cell positions measuredY J 7-37 (Sffl 7-9Control rods: Z1,Z2- safety? C1,C2,C3,C4- compensation?

R- regulation.

11,/ cm

Fig.2- Fuel element and ending structures

1- fuel element, 2- upper end,3- Uranium in 3 concentric layers,4- lower end and 2 support plates.

156

of 3 concentric layers with uranium of 36% enrichment having 60cm length and0.7mm thickness, cladded with 0.9 mm thick aluminum tubings of hexagonal externalform and circular inner forms. Uranium average weight in each fuel element is 40.2g.In the core there are also :

- An additional reflector made of beryllium juxtaposed to the old graphite reflector,

A central neutron trap with beryllium surrounding and full of water,Two pneumatic irradiation channels (dry channels) at positions 7-1 and 13-2 andone 'wet" irradiation channel at position 1-4,

7 control rods, all moving in water filled aluminium guide tubes (2 safety rods AZand 4 compensation rods KC, 65cm long and 27mm outer diameter, made ofboron carbide, together with one regulation rod AR of stainless steel).

Fuel elements are inserted into 2 support plates of 2cm and 4cm thicknessrespectively at their lower ends. At their upper ends they are juxtaposed by theirhexagonal heads (Fig. 2). The upper and lower parts of the core are then composedof water and aluminum of 11.7cm and 14.8cm thickness respectively.

3, AXIAL DISTRIBUTION OF THERMAL NEUTRON

Thermal neutron relative distributions in the z-direction of the core are measuredby foil activation method at different positions, particularly at the position 1-4 of thealready existing wet channel. Other positions are selected for their marked influenceof control rods on the neutron distribution. For these positions, the correspondingfuel element must be withdrawn in order to insert activation foils. In these cases itwas shown [1] that thermal neutron distribution at the water column created by fuelelement withdrawal coincided with the one at the surface of fuel element at the samecoordinate.

For experimental realization, 9mm x 2mm x 0.1mm copper foils of 8 ing weightare irradiated along plexiglass plate put in the center of the created water column.This practice of average approximation is justified by the small thickness of uraniumconcentric layers separated by relatively large water interstics; it is also justified byexperimental data obtained from microflux measurement carried out inside the fuelassembly at the first start-up of the reactor. The foil irradiation time is 15 minutes at2.5 kW and the cooling time is about 24 h. The gamma activity of the foil (511 KeVand 1346 KeV peaks) is determined by using 70cm" high purity Ge detectorconnected to a Canberra MCA, the beta activity is also measured by beta counter.

During irradiation, the two safety rods AZ are completely withdrawn. A numberof thermal neutron distributions at 3 representative positions, i.e 7-3 and 7-9 (cellsinside the core) and 1-4 (wet channel at the core limit), has been measured fordifferent configurations of remaining KC and AR rods in order to show the influenceof nearby inserted control rods. (Dominant reactivity worth is from the 4 KC rods,about 3$ each, while the 0.5$ AR rod is of less influence).

Fig. 3 represents some experimental data on thermal neutron distributions, originof the z-axis being taken at the upper limit of the core as already used for indicationof control rod position. One can see that not all the experimental curves aresymmetrical, the neutron distributions are deviated toward the lower part of the core,depending on the inserted depth of control rods. This nonsymmetrical featureappears clearly from the values zmax of the ordinate at which the distribution attainsthe maximum value. It can be seen for each core position there can exist a control

157

4J

1—4

QJ

0.5

0.5

I.O

©•5

a

Position 7-3

2Position 7-9 O

> a+

4-

Position 1-4

10 20 30 40 50 H(em)

Fig- 3 - Axial distributions o-f the thermal neutronsat the positions 7-3, 7-9 and 1—4 (error ±3%)# No 1, * No 2, H No 3,O No 4, x No 5, 4 No 6,f No 7, n No 8, V No 9, * No 10.

158

rod configuration that gives almost symmetrical distribution curve about the middleplane of the core (zmax = 30cm). For different distributions, zmax values are given inTable 1 together with other parameters deduced from experimental data.

Table 1. Parameters obtained from neutron distribution measurementat core positions 7-3, 7-9 and 1-4.

Corepos.

7-3

7-9

1-4

No.

1

23

45

678

9

10

Inserted depth, of control rodKC113.242.0

35.0

65.000.0

00.065.000.000.035.0

KC265.0M.O

35.0

19.539.0

39.065.065.000.035.0

KC365.0

65.0

35.0

19.539.0

39.011.543.048.535.0

KC413.242.0

35.065.0

65.0

65.011.543.048.035.0

AR65.0

36.5

31.500.0

00.0

31.065.033.021.531.5

(cm)

30.0

33.4

35.630.030.8

31.831.8

34.335.335.7

k ±0.01

1.30

1.32

1.32

1.301.30

.31

.32

.33

.33

.33

K(m"2)

16.7 ± .4

16.6 ± .4

15.8 ± .316.8 ± .4

16.6 ± .2

16.7 ± .317.1 ± .217.0 ± .316.8 ± .2

16.3 ± .2

Hcrr(cm)

76.8 ± .9

77.1 ± .9

79.1 ± .9

76.6 ± .8

77.2 ± .5

76.9 ± .675.9 ± .476.2 ± .776.6 ± .577.8 ± .4

Note: Control rod position in cm, 65 - fully inserted, 0 - completely withdrawn.

From Table 1, there are only two symmetrical distributions associated withpositions 7-3 and 7-9, i.e cases No 1 and No 4 respectively. These symmetrical casescorrespond to the configurations where the 2 nearest KC rods (KC2, KC3 and KC1KC4 respectively) are fully inserted while the two other KC rods are almostcompletely withdrawn. But for the external position 1-4, an analogous configurationof KC rods does not give rise to symmetrical distribution (case No 7: the nearestKC1, KC2 rods fully inserted, zmax = 31.8cm). For the other cases, the dissymmetryis increased (zmax higher) when the rods rearest to the studied position are eithercompletely withdrawn or completely inserted (except cases Nos 3 and 10), and thisfeature is accentuated with increasing inserted depth of the remaining rods. For casesNo 3 and No 10 with all 4 KC rods inserted at intermediate depth of 35cm, thedissymmetry is particularly the greatest.

The dissymmetry of axial distribution curves can be best expressed analyticallyby [2]

<I>(z) = [A + C (z-30)] cos B7. (z-30) (1)

where A, C, Bz are constants, and C = 0 for symmetrical distributions, Bz is the axialbuckling.

The parameters A, C, Bz are determined from experimental data by least squarefit [3]. From (1) it is deduced the inhomogeneous coefficient of neutron field on thez-directinn defined by

kz = <I>max / (I>av (2)

where <I>max = cI»(zmax) , maximum value of <I>, and <I>av its average given by

159

<I>av = (I/H) <I>(z)dz (3)

with H = 60cm.

The menn value of kz at position 1-4 is 1.33 ± 0.01, and at positions 7-3 and 7-9it is 1.31 ± 0.01. The global mean value of kz calculated from these experimental datais

k7. = 1.32 ± 0.01

Values of zmax and kz are also used as parameters guiding the choice of sampleirradiation position inside the reactor core (for example, irradiation of silicon bar).

4. CORE EFFECTIVE HEIGHT

The distribution function does not equal zero at physical limits of the core. Bydefining the effective height Hcff of the core as the distance between the positionswhere <I>(z) = 0, i.e by putting

Heff = H + Au + A| (4)

where Au and A| are'respectively the extrapolated distances at the upper and lowerboundary of the core determined by the conditions

<I>(-AU) = <n(H + A|) = 0 (5)

The conditions (5) applied to function (1) give rise to two solutions, i.e the zerosof the two factors <I>i = A + C(z-30) and <I>2 = cos Bz (z-30) . From the fitted valuesof A, C, Bz, it is seen that A > 0, C < 0 and the zero of <I>i is much greater than thatof <I>2 for all the cases studied. In consequence, one must take the conditions (5) as

= 0 (6)

As <I>2 is symmetrical function, then Au = A| = A for all symmetrical or nonsymmetrical distributions, and (6) becomes

cos(± BzxHCff/2) = 0 (7)

where Heff = H -f 2A. It turns out from (7) that Bz is the buckling of the reactor inthe z-direction as already noted [4]. Experimental values of Bz are shown in Table 1for each core configuration. The deduced average is

Bz2 = (16.6 ± 0.6) m"2

The expression of Heir calculated from (7) is

Hcff = n / Bz (8)

Experimental values of Heff are given in Table 1. The resulted average values ofeffective height and extrapolated distance are

Hcff = (77.0 ± 0.9) cm

and A - (8.5 ± 0.5) cm.

The precisions involved in the above results are given by least square fit.

5. CONCLUSION

160

From the measurement of thermal neutron relative distributions on verticaldirection at different positions and control rod configurations in the core of the Dalatnuclear research reactor, it has been deduced several core parameters:

The inhomogeneous coefficient of neutron field in the vertical direction, kz =1.32 ± 0.01,

- The axial buckling, Bz2 = (16.6 ± 0.6) m"2 ,

The axial extrapolated distance, A = (8.5 ± 0.5) cm ,

The effective height of the core, Heff = (77.0 ± 0.9) cm.

ACKNOWLEDGEMENTS

The authors are indebted to Prof. Pham Duy Hien, Drs. Tran Ha Anh, NguyenTac Anh, Pham Ngoc Chuong, Tran Khanh Mai, Engineers Ngo Phu Khang, PhamVan Lam, Huynh Dong Phuong for discussion, correction and manuscript reading.It is pleasure to acknowledge the efforts of the reactor staff, and the equipmentassistance from Nuclear Physics Department and Radoprotection Department.

REFERENCES

1. Galanin A.D. - Theory of heterogeneous reactor, Moskow, Atomizdat, 1971(in Russian).

2. Tong L.S. and Weiman J. - Thermal analysis of PWR, ANS 1970.

3. Szatmary Z. - Data evaluation problems in reactor physics. Theory of programRFIT. Rep. KFKI-1977-43, Budapest.

4. Murray R.L. - Nuclear reactor physics, Prentice Hall, 1959.

161 I E:'I le

Proceedings o f the 4th National Conference on Physics I 111 111 IIIi mil HUM n IL

Hanoi, Vietnam, 5 - 8 October 1993. VN9706b ib

THE RADIAL DISTRIBUTION OF THE NEUTRON FIELDIN THE CORE OF THE DALAT REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Hochiminh City, Vietnam

Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh

Nuclear Research Institute, Dalat, Vietnam

ABSTRACT

Determination of the radial distribution of the thermal neutron field in the coreof the Dalat reactor is done by the Cu foil activation method. The measured data arefitted by the least square method to determine several physical parameters of thereactor, as follows :

1. Buckling B r2 - (84.6 ± 5.5) 10"4/cm2

2. The effective radius Rgff = (27.6 ± 1.0) cm3. The extrapolation distance X = (8.7 ± 1.0) cm4. The unequal coefficient of the effective multiplication Kr = 1.77 ± 0 . 1 1

I. INTRODUCTION.

The Dalat reactor was reconstructed from the former reactor TRIGA Mark II, inwhich the new core was surrounded by the old graphite reflector and an additionalberyllium reflector layer. The complicated structure of the core with a small size leadsto a strong variation of the radial distribution of the neutron field. Determination ofthis distribution is necessary to exploit the reactor as well as supply data for calculatingparameters : the buckling, the extrapolation distance, the unequal coefficient of theradial neutron field and the effective radius of the core.

n . STRUCTURE OF THE CORE.

The reactor core in the cylindrical shape of 43.6 cm diameter (including theberyllium reflector layer) is surrounded by the graphite reflector of 34.6 cm thickness(Fig. 1). The core is loaded with 89 fuel elements of Russian-made WR-M2 type . Afuel element has an overall height of 86.5 cm including 60 cm aluminum-uranium fuelpart of 36% uranium enrichment. The fuel element consists of three coaxial tubes, eachof which contains an 0.7 mm fuel layer wrapped in two aluminum layers of 0.9 mmthickness (Fig.2). The average weight of U2-*-> in each fresh fuel element is 40.3 g.

163

Cell I-1

Fig. 1 Reactor core arrangement1. Fuel element ; 2. Beryllium reflector 3. Neutrontrap ; 4. Wet channel ; 5. Pneumatic transfer tube ;6. Safety rod ; 7. Shim rod ; 8. Regulating rod ;9. Graphite reflector ; Cell 1-1 : The first cell of thecore, cell i-j denotes the cell at i1'1 row and j 1 ' 1 column.

l.S

0.? 0.7 03

LI,•1 >

Al

Fig. 2. Cross section of the fuel element( Unit in nun )

Fig. 3 Cross section of the plexiglass rod

164

In addition, in the core there are a neutron trap , six boron-carbide absorptionrods (two safety rods AZj, AZ2 and four shim rods KCi-^KC4), an automaticregulating rod AR of stainless steel, a "wet" irradiation channel at cell 1-4, twopneumatic transfer channels at cells 7-1 and 13-2. The neutron trap is a beryllium blockoccupying 7 central cells and having a water column inside of 65 mm diameter. Crosssection area of the active core (composing of 89 fuel elements, seven control rods,three irradiation channels, a neutron trap, exclusive of the beryllium reflector layer) is1124.7 cm^ . So the equivalent radius of the active core is R = 18.9 cm

The reactor has a thermal column and four horizontal tubes. The 1st , 3 r" , 4 t n

tubes have the void extension parts in the graphite reflector.

ITL THE RADIAL DISTRIBUTION OF THE THERMAL NEUTRONFIELD.

The thermal neutron field is determined by activating Cu foils of 9 mm x 2 mmx 0.1 mm size and (5 -s- 8) ing weight at 2.5 kW in 15 minutes . A solid plexiglassrod of the same size as a fuel element is made to allocate foils (Fig.3) and is placed inthe measured position instead of the fuel element. Measured positions are composedof two groups of cells in the core and are chosen far from the absorption rods and theirradiation channels. The first group is of cells 4-1 , 5-2 , 5-3 , 6-3 , 6-4 , 6-5 and thesecond - cells 6-8 , 6-9 , 6 - 1 0 , 5 - 1 0 , 4 - 1 0 (Fig. 1). These groups of cells have radialdirections through the solid parts of the graphite reflector. It allows the radialdistribution of the neutron field to be characteristic, not being affected by the strongabsorption or diffusion materials. Calibration of the reactor power is carried out bybeta counts of a Cu foil placed at an irradiation water hole in the rotary specimen rackof the graphite reflector.

Table 1 presents the results obtained. For the considered positions of equalapproximation radii the presented experimental values are averaged. The measuredvalues at several positions close to the boron-carbide rod KC2 are not used either.From Table 1 it is evident that the experimental results in the two groups of cellscoincide in error range (exclusive of point number 2 with radius R = 5.2 cm , the tworesults are greatly different).

Figure 4 illustrates the radial distribution of the thermal neutron field for averagevalues of the two measuring groups. It is seen from Figure 4 that the thermal neutronfield has maximal value at close to the neutron trap, primarily decreases quickly andfinally increases at the fuel elements near the beryllium reflector. The unequalcoefficient of the radial distribution of the thermal neutron field is :

Kr = < I ) m a x / < ( I ) > = 1 - 7 7 ± o n

where the maximal value of the thermal neutron field is ^max = 1-&7 ± 0.10 and theaverage value is « t » = 1.05 ± 0.03.

165

I n t h e c o r e c e n t r e ( r - 0 ) : <$> - 3 , 3 8 ± 0 , 0 4

18 R (cm)

Figure d. The radial distribution.- of the thermal

neutron field.

according to expression (l)

- - - - - extension part of function :

&-_Jo(a2(r - a )

166

Table 1. The radial distribution of the thermal neutron field .

No

010203040506070809101112131415161718192021222324 •252627282930

R ( ± .05) cm

4.35.2

5.556.97.27.6

8.258.79.09.9

10.4511.0511.4512.313.113.313.613.814.114.315.5515.916.2516.5517.0

17.2517.6519.019.319.9

TTie neutronThe 1st group1.93 ± 0.051.56 ± 0.041.47 ± 0.051.19 ±0.061.06 ±0.051.18 ± 0.171.11 ±0.04

1.13 ±0.040.99 ± 0.021.04 ± 0.021.00 ±0.021.01 ± 0.020.94 ± 0.030.93 ± 0.03

0.93 ± 0.030.89 ± 0.06

0.83 ± 0.02

0.93 ± 0.020.92 ± 0.040.87 ± 0.030.94 ± 0.040.95 ± 0.041.06 ± 0.041.08 ± 0.04

flux (relativeThe 2nd group

1.80 + 0.061.90 ± 0.051.60 ±0.071.31 ±0.071.16 ± 0.051.09 ± 0.041.04 ± 0.051.10 ±0.04

1.04 ± 0.031.07 ± 0.040.98 ± 0.020.97 ± 0.030.99 ± 0.03

0.92 ± 0.020.87 ± 0.030.93 ± 0.03

0.91 ± 0.020.86 ± 0.05

0.85 ± 0.070.92 ± 0.020.88 ± 0.030.92 ± 0.031.01 ± 0.041.01 ± 0.011.04 ± 0.021,05 ± 0.02

unit)Average

1.87 ±0.101.73 ± 0.241.54 ±0.101.25 ±0.101.11 ± 0.081.13 ±0.091.08 ± 0.061.10 ± 0.041.13 ± 0.041.01± 0.041.06 ± 0.030.99 ± 0.020.99 ± 0.030.97 ± 0.040.93 ± 0.030.92 ± 0.020.87 ± 0.030.93 ± 0.040.89 ± 0.060.91 ± 0.020.86 ± 0.050.83 ± 0.020.85 ± 0.070.92 ± 0.020.90 ± 0.030.90 ± 0.040.98 ± 0.060.98 ± 0.051.04 ± 0.021.07 ± 0.03

IV.CORE.

DETERMINATION OF THE PHYSICAL PARAMETERS OF THE

The radial distribution of the thermal neutron field may be described by the two-group neutron diffusion theory for the reactor with reflector. To avoid the influence ofthe neutron trap, the radial distribution function is calculated from the radius in adiffusion length L = 2.8 cm far from the neutron trap . Then, according to [1] :

(r - a3)] + a4 - a3)] (1)

where spherical Bessel function JQ describes the thermal neutron field distribution inthe core in a diffusion length far from the beryllium reflector, Bessel function IQ

167

describes the influence of the beryllium reflector on the neutron field near the reflector.Constants a] , a2, a3 , a^ and 35 are determined from experimental data by the leastsquare fitting method [2J.

The calculation results give :

<i| - 1.18 ± 0.02 , a2 = 0.092 ± 0.003 , a3 - 1.4 ± 0.5

a4 - 0.0035 ± 0.0002 , a5 = 0.39 ± 0.03

The fitting curve in Figure 4 describes well the experimental data.

Function JQ describes the neutron field distribution in framework of the one-group neutron diffusion theory for the bare reactor . So radial buckling of the coreequals

B ]2 = a 22 « (8 4 6 ± 55) 10"4/cm2 (2)

The effective radius of the core is determined from boundary condition:

JoPMReff - a 3 ) ] = 0 (3)

Hence,

Reff - 2.405/Br + a 3 = (27.6 ± 1.0) cm (4)

Because the equivalent radius of the core is R = 18.9 cm, the radial extrapolationdistance is :

^r = Reff-R = (8-7 ± 10) cm (5)

V. ACKNOWLEDGMENTS.

The authors are grateful to Professor Pham Duy Hien, Dr. Trail Ha Anh, Dr.Nguyen Tac Anh, Professor Cao Chi, Dr. Pham Ngoc Chuong, Dr. Nguyen PhuocLan, Dr. Tran Khanh Mai, Dr. Nguyen Tien Nguyen for their helpful comments. Theauthors also thank the reactor operation shifts for helping perform the experiments andthank Nuclear Physics Department, Radiation Safety Department for makingconditions comfortable to use experiment equipment.

VI. REFERENCES

1. R. L. MURRAY,Nuclear Reactor Physics, Prentice-Hall, 1957.

2. Z. SZATMARY,Data Evaluation Problems in Reactor Physics, Theory of Program RFITRep. KFKI - 1977-43, Central Research Institute for Physics, Budapest.

168

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993.

VN9700011

DISTRIBUTION OF THE THERMAL NEUTRON FIELDAROUND THE GRAPHITE REFLECTOR

OF THE DALAT NEUCLEAR RESEARCH REACTOR

Ngo Quang HuyCenter of Nuclear Techniques, Ho Chi Minh City, Vietnam.

Ha Van Thong, Vu Hai Long, Ngo Phu Khang,Nguyen Due Binh, Nguyen Minh Tuan, Le Vinh Vinh.

Nuclear Research Institue, Dalat, Vietnam.

ABSTRACT

Thermal neutron flux distributions around vthe graphite reflector of the Dalatnuclear research reactor are determined by the method for neutron activating Cufoils. The major results are as follows:

a. The axial distributions at the inner and outer margins of the graphite reflectorhave dissymmetrical shapes, similar to axial distributions in the core. There is adissimilarity between the distribution curves at the inner margin and those at theouter margin of the reflector.

b. The radial distribution on the upper surface of the graphite reflector ismeasured and is described by the two-group neutron diffusion theory. The maximalvalue of the curve lies at the position of Rmax = 22.5 cm.

c. The distribution in the twenty water irradiation holes around the rotaryspecimen rack is obtained.

1. INTRODUCTION

The Dalat nuclear research reactor was reconstructed from the former TRIGAMARK II reactor. The new core is surrounded by the old graphite reflector and anew additional berylliuin layer. The radial distribution of the thermal neutrons in thenarrow active core is strongly varied with radius [1] and the axial distributions aredissymmetrical relative to the center of the core [2]. The determination of theneutron flux distributions on the surface of the graphite reflector and around therotary specimen rack is useful for reactor exploitation and reactor calculation.

2. REACTOR CORE AND GRAPHITE REFLECTOR

The reactor core of a cylinder shape with 60 cm height and 43.6 cm diameterincluding berylliuin layer is surrounded by the old graphite reflector (Fig. 1). Thereactor is loaded with 89 fuel elements of WR-M2 type, a neutron trap, 7control rods and 3 irradiation facilities. A fuel element has an overall height of 86.5

169

r c it

• i .1 .'•«; i ( ) iM i y '(-

'• ••••••[plsj?.

/

Fig. 1 Reactor core arrangementI. Fuel clement ; 2. Beryllium reflector 3. Neutrontrap , 4. Wet channel ; 5. Pneumatic transfer tube ;6 Safety rod ; 7. Shim rod , S. Regulating rod ;9. Graphite reflector ; Cell 1-1 : The first cell of thecore, cell i-j denotes the cell at i t n row and j l n column.

Extracting well

Position 5-1(or 9-11)

Active core

Holder of Cu foils

Rotary specimen rack

F'9-J2 Elevation of the graphite reflector and positions of activating foils.

170

CD

01

2.0

0 1O ri (err.)

Fig. 3 Axial distributions of the thermal neutrons atthe inner (•) and outer (A) margins of the graphite reflector(error of 5%).

171

cm including 60 cm uranium-aluminum fuel part (36% enriched U235). The equivalentradius of the active core ( fuel part of the core) is 18.9 cm.

The old graphite reflector is a rim block, having 34.6 cm thickness (35.5 cmwith aluminum cladding) and 45.2 cm inner diameter. Inside the graphite block thereare three big holes coaxial with three horizontal beam tubes. A deep drain of 30 cmdepth and 7 cm width is made on the upper part of the graphite reflector forinstalling the rotary specimen rack which has 40 irradiation holes with diameter of3 cm and depth of 28 cm in distance of 35 cm from the active core center. Therotary specimen rack operates completely in water.

The beryllium reflector of an average thickness of 2.9 cm consists of the rodshaving the same outer shape as a fuel element and the beryllium rings formed bysome serrated blocks.and placed just inside the graphite reflector.

3. AXIAL DISTRIBUTIONS OF THERMAL NEUTRONSAND OUTER MARGINS OF THE GRAPHITE REFLECTOR

AT INNER

Determination of the neutron flux distributions inside the graphite reflector isunable because of its solid structure. Therefore measurements are carried out at theinner and outer margins of the reflector (Fig.2). The Cu foils of 9 mm x 2 mm x0.1mm size and of about 5 mg mass are irradiated in 15 minutes at 2.5 kW power.Fig. 3 presents the axial distributions of the thermal neutrons at the positions 5-1 and9-1 1 at the inner margin and corresponding positions at the outer margin of thereflector. It is obvious from Fig. 3 that the axial distribution curves havedissymmetrical shapes, similar to those in the active core [2], and are expressed by thefollowing formula :

O(z) = [A + C(z-30)] cos[B(z-30)] (1)

where A, B, C are constants. The constants Ajn, Bjn, Cjn and A o ut , B o ut , Co ut aredetermined by the least square fitting method for the inner and outer marginsrespectively. Calculation results are presented in Table 1.

Table I. Fitting data for Bjn , B o u t , Ajn /Ao u t and Cj n /C o u t .

Parameters

Bin

Bout

Ain/Aout

^iiv^out

5-

0.038

0.037

2.43

2.06

1

±

±

±

±

cell

0.001

0.001

0.05

0.02

9-

2.54

2.28

11

±

±

cell

0.04

0.01

172

Table 1 shows that Bjn equals B o u t while Ajn/A0U( ratio is not equal toCjn/Cout ratio. It means that the curves at the inner margin are not similar tothose at the outer margin. This dissimilarity may be explained by the influence ofthe rotary specimen rack and irradiation holes placed on the upper part of thegraphite reflector. However the dissimilarity is not significant because the differenceof Ajn /Ao u t and Cjn/Cout is only 12.5%.

4. THERMAL NEUTRON FIELD DISTRIBUTION ON THE UPPERSURFACE OF THE GRAPHITE REFLECTOR

The thermal neutron field distribution on the upper surface of the graphitereflector is only measured from the irradiation hole position to the outer marginbecause of the existence of the extracting well over the reflector (Fig. 2). Theexperimental result (Fig.4) shows a rapid decrease of the neutron fluxes with radius.

The experimental data on the thermal neutron fluxes are expressed by the two-group diffusion neutron theory as follows [3]:

F(r) - a i Ko(Xlr) + a2 I0( xjr) + a3 KQ( x2r) + a4 Io( x2r) (2)

where aj - 34 are constants,

x ] 2 = 1 /x + (Ti/Hgff)2 ; x 22 = (1/L)2 + (7i/Heff)

2 (3)

T : neutron age,L : diffusion length of thermal neutrons.Heflf = (77.0 + 0.9) cm : effective height of the core [2]

The constants a] - a4 are determined from experimental data by the least squarefitting method with T and L parameters chosen for the two media : graphite andwater. Calculation results indicate that the graphite medium ( x 2 =0.9) is moresuitable than the water medium (x 2 = 367).

Fig.4 presents the calculation curve that increases with radius, attains maximumvalue at the position of R m a x = 22.5 cm and after that decreases. So the maximumof the neutron flux distribution is not lain at the irradiation water hole. It may beexplained by the existence of the additional beryllium reflector.

5. THERMAL NEUTRON FIELD DISTRIBUTION AROUND THEROTARY SPECIMEN RACK.

Distribution of the thermal neutrons around the rotary specimen rack isdetermined by activation of 20 Cu foils placed in 20 irradiation water holes at thedepth of 15 cm. Fig. 5 shows the experimental result of the relative neutron fluxesd>/<0>, where <O> is the average value of 20 measured data. From Fig. 5 it is seenthat O/<O> is greater than 1 at the 3rc* and 4 t n horizontal beam tubes and is less than1 at the 1st and 2n(* beam tubes. The maximum and minimum values are 1.13 ±0.04 and 0.89 ± 0.03 respectively. Dissymmetry of the distribution is 1.27.

173

O.J

0,2

N

1 rrad i M(ionliul e

i

N

: X(«-)

Fig. 4. Radial distribution of the thermal neutrons onthe upper surface of the graphite reflector.

• : Experimental data.— : Function (2).

FicLJl Thermal neutron fluxes around the 20 irradiation holesof the rotary specimen rack.

• 1.02 ± 0.03: Irradiation hole with Cu foil andrelative flux value at the hole.

o : Irradiation hole without any Cu foil.

174

6. ACKNOWLEDGMENTS

The authors are indebted to Prof Pham Duy Hien, Dr. Tran Ha Anh, Dr.Nguyen Tac Anh, Prof. Cao Chi, Dr. Pham Ngoc Chuong, Dr. Nguyen Phuoc Lan,Dr. Tran Khanh Mai, Dr. Nguyen Tien Nguyen for interest in the work anddiscussion. It is a pleasure to thank the efforts of the reactor staff and the equipmentassistance from Nuclear Physics Department and Radiation Protection Department.

7. REFERENCES

1. Nguyen Due Binh, Ngo Quang Huy, Ngo Phu Khang, Vu Hai Long, Ha VanThong, Nguyen Minh Tuan, Le Vinh Vinh.

"The radial distribution of the thermal neutron field in the core of the Dalatnuclear research reactor".

Preprint of the Nuclear Research Institute, Dalat, Vietnam, E 091 - 91, 1991.

2. Nguyen Due Binh, Ngo Quang Huy, Vu Hai Long, Ha Van Thong, NguyenMinh Tuan, Le Ba Vien, Le Vinh Vinh, Perez Martin David, Garcia Yip Fernando.

"Effective height of the core of the Dalat nuclear research reactor".Preprint of the Nuclear Research Institute, Dalat, Vietnam, E 035 - 89, 1989.

3. R.L. Murray. "Nuclear reactor physics".Englewood Cliffs, N.J. Prentic Hall, Inc, 1957.

175

._„., ,„_,, _„,„ mmin ••!•! miB |

VN9700012Proceedings of the 4th National Conference on Physics

Hanoi, Vietnam, 5 - 8 October 1993.

CALCULATION OF FUEL BURN-UP AND FUEL RELOADING

FOR THE DALAT NUCLEAR RESEARCH REACTOR.

Nguyen Phuoc Lan ,Ngo Quang Huy.Center of Nuclear Techniques, Ho Chi Minh City, Vietnam.

Ha Van Thong , Do Quang Binh.Nuclear Research Institute, Dalat, Vietnam.

ABSTRACT

Calculation of fuel burnup and fuel reloading for the Dalat nuclear researchreactor was carried out by using a new programme named HEXA-BURNUP,realized in a PC computer. The programme is a modification of the HEXAprogramme, based on the two-dimensional three-group neutron diffusion theory,with the introduction of the effects of control rods inserted at different depths andnuclide concentration variations. The HEXA-BURNUP programme is used tocalculate the following parameters of the Dalat reactor:

a. Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92fuel elements. The effective multiplication coefficients equal 1 within the errorranges of less than 0.38%.

b. The thermal neutron flux distribution in the reactor. The calculated resultsagree with the experimental data measured at 11 typical positions.

c. The average fuel burn-up for the period from February 1984 to September1992. The difference between calculation and experiment is only about 1.9%.

d. 10 fuel reloading versions are caculated, from which an optimal version isproposed.

I.INTRODUCTION

The Dalat nuclear research reactor was reconstructed from the former TRIGAMARK II reactor built in 1963 with the nominal power of 250 kW. The newlyreconstructed reactor reached the first criticality on the 1st of November 1983 andits planned nominal power of 500 kW in February 1984.

177

Il<«v«.tion of reacto • system.

1. Core2. r.r-«|ihttc Rcflceto •) . Suehinij Tub<y4. l'rl"ftW Ooolinr. S.otc«5. Adriltional Shield .nr,.

t ic Ir.-tnsf ovlliolron tmr

178

The Dalat reactor is of swimming pool type loaded with hexagonal fuel elementsof Russian-made VVR-M2 type. The design calculation for the reactor was carriedout by using the two-group neutron diffusion theory and the multi-group approxima-tion [1]. In the HEXA programme [2] the two-dimensional three-group neutrondiffusion theory is used to calculate the neutron tlux distribution for reactors similarto the Dalat reactor. However the calculation models presented in [1] and [2] do notinclude the effect of intermediate depth of the control rods in the real arrangementof the reactor core. This important effect is introduced into the programme modifiedfrom the HEXA programme. A combination of this modified programme and theprogramme calculating the nuclide concentration variations [3] leads to a newprogramme named HEXA-BURNUP . The HEXA-BURNUP programme, realizedin a PC computer, is used to calculate the parameters of the critical core configura-tion, the neutron flux distribution, the average fuel burn-up, the fuel burn-updistribution and the fuel reloading version for the Dalat reactor.

2.REACTOR DESCRIPTION

Fig. 1 presents a view of the Dalat reactor, of which the new components consistof the reactor core, the extracting well and the cooling system. The undismountedcomponents of the former reactor are the reactor tank, the concrete shielding, thegraphite reflector, the beam tubes and the thermal column.

The reactor core has a cylinder shape, 60cm in height, 43.6cm in diameter andvolume of about 100 litres. Inside the core are placed 89 fuel elements, 7 control rods(6 boron carbide rods and 1 stainless steel rod), a neutron trap, 3 irradiation channels(1 wet channel and 2 pneumatic transfer channels) and an additional berylliumreflector (Fig.2). All the elements are fixed by two perforated plates at the bottomof the core.

The fuel elements used in the new reactor belong to the Russian-made VVR-M2type (Fig.3). They are made of uranium-aluminum alloy with aluminum cladding. Onaverage each fuel element contains about 40.2g of U * distributed on three coaxialfuel shells, the outermost shell has the shape of hexagonal cylinder with a size of32mm, the two inner shells have cylinder shape with diameter of 22mm and 11mmrespectively. Each shell consists of three layers, the fuel layer with a thickness of0.7mm is wrapped between two alluminum alloy clading layers of 0.9mm thickness.The gaps of (2.5-3.0)mm between the shells of a fuel element are used as thepassages for water circulation. The total length of the fuel element is 86.5cm ,ofwhich the fuel containing part is 60cm, the remainder parts are made of aluminum.

The old graphite reflector surrounding the core is a rim block with a thicknessof 34.6cm and an inner diameter of 45.2cm. Inside the graphite block there are threebig holes coaxial with three horizontal beam tubes. A deep drain of 30cm in depthand 7cm in width is made on the upper part of the graphite reflector for installingthe rotary specimen rack which has 40 irradiation holes with 3cm diameter andoperates completely in water. The reflection material above and under the coreconsists of aluminum and water.

179

22

35

ul-'uol element of VVK-M2 type

180

Beryllium material consists of the rods having the same outer shape as fuelelements and beryllium rings formed by some serrated blocks and placed just insidethe graphite reflector.

3. HEXA-BURNUP PROGRAMME

3.1 Calculation model

The structure of the reactor core and the fuel elements as described in Section 2allows the core to be composed of the homogeneous hexagonal cells with charac-teristics averaged over a cell [4]. The effect of the close arrangement of the fuelelements will be taken into account by introducing the mass effect in the thermalenergy range [4] and the resonance escape probability in the resonance energy range[5].

All the cells in the core, except the control rods, have the same length. Thereforethe neutron physics characteristics of the reactor may be determined for a horizontalcross-section of the core. It allows the use of two-dimensional geometry with thehexagonal cells in the calculation model. For this model a three-group neutrondiffusion theory is available because the metal/total material ratio in a cell is 0.448falling within (0.3-0.66) ratio range [6]. The first group consists of the fission neutronswith (5KeV - 1 OMeV) energy, the second group - the resonance neutrons with (0.2eV- 5KeV) energy and the third one - the (0 - 0.2)eV thermal neutrons.

Neutron fluxes in the reactor core are determined by the system of 3 equations:

3-vXD| VNM) + Zi<I>i = (1/keff) 2 v 2 fi *i

/ = 1

= Lsll <I>1 (1)

=<p ISI2 <I>2

where i-index indicates the ith group (i= 1,2,3), i is the neutron flux, Dj is the diffusioncoefficient, 2j is the cross-section of releasing neutron, 2n is the fission cross-section,Ssii is the slowing-down cross-section, v is the number of neutrons produced by afission, <p is the resonance absorption escape probability, kcff is the effective multi-plication coefficient.

3.2 The control rod effect.

The presence of the control rods inside the reactor core has a significant influenceon the neutron flux distribution. Furthermore this distribution depends on thenumber and the depths of the inserted control rods. In order to use the calculationmodel of two-dimensional geometry it is necessary to take into account the controlrod effect for the practical working conditions of the reactor core.

181

It is proposed that in the case of the reactor core with control rods withdrawncompletely from the core or inserted fully inside the core, the axial neutron fluxdistribution should belong to the cosine form. When the control rods are inserted atintermediate depths, the axial distribution has a modified cosine form, of which theparameters are determined from the continuous condition of the neutron fluxdistribution. This modified cosine function is used to average the boundary conditionat the control rod surface. So an effective boundary condition is obtained for thetwo-dimensional geometry model.

Neutron fluxes belong to the following conditions at the outer boundary of thereactor :

2 D i(/Tv* >I>i + 'I>i = 0 (2)

and at the surface of the control rod cell :

yj(/lV)<l>j + <[>j = 0 (3)

where yi are effective values averaged over the axial distribution of neutron fluxes :

H Hy i = / y i (z) <I> i (z) dz / / <l> i (z) dz (4)

n o

3.3. HEXA-BURNUP programme

The system of 3 equations (1) is solved with the boundary conditions (2) and (3)for the Dalat reactor by using a progamme modified from the HEXA programme.

A combination of the modified HEXA programme calculating the neutron fluxesfor all the cells and the programme calculating the nuclide concentration variationfor each cell [3] leads to a complete programme named HEXA-BURNUP. Thisprogramme, realized on a PC computer, permits the calculation of the reactorphysics parameters for the real core configurations.

4. CALCULATION RESULTS

4.1. Pnrnmeters of the critical core configurations

Table 1 shows the results calculated for the parameters of the critical coreconfigurations in physics start-up, energy start-up and working conditions. FromTable 1 it is obvious that for all the critical situations, in which the depths of thecontrol rods are given from experimental data, the effective multiplication coeffi-cients keff calculated equal 1 within the error range of less than 0.38%.

182

Table 1. Parameters of the critical core configurations.

Number offuel elements

|7)

69

72

74

86

88

89

92

Crilical U235 mass (g)

Experiment{71

2781.8

2897.4

2977.1

3461.8

3537.0

Calculation

2775.11

2895.77

2976.21

3458.83

3539.27

3548.88

3690.50

Depth of shimrods (mm)

(71

KC4 = 6

KC4 = 195

KC2 = 442

4KC « 363

4KC = 462

4KC = 400

4KC = 470

kcff

0.9978

1.0007

1.0028

1.0000

1.0002

1.0038

1.0015

Excessreactivity

(%)

0.24

1.84

6.04

7.58

7.14

7.90

4.2. Thermal neutron fluxes

Table 2 presents the calculation values and experimental data [8] for the thermalneutron fluxes at several positions in the core. Theoretical results well agree withexperimental data at 10 from 11 measured positions. In the II1 position, i.e. inboundary of the core the difference between theory and experiment is 5%.

Table 2. Thermal neutron fluxes in June 1989.

Fuel element

6-5

6-4

6-3

5-3

5-2

4-1

6-8

6-9

6-10

5-10

4-10

Thermal neutron fluxes

Calculation

(1013n/cm2/sl

0.52

0.37

0.33

0.28

0.29

0.30

0.51

0.35

0.31

0.27

0.28

Relative

1.50

1.05

0.96

0.80

0.84

0.85

1.48

1.02

0.90

0.79

0.80

Experiment (8](Relative)

1.33 ±0.32

1.06 ±0.11

0.92 ± 0.04

0.89 ±0.10

0.87 ± 0.06

0.94 ± 0.08

1.39 ±0.32

0.99 ± 0.03

0.89 ± 0.03

0.80 ± 0.04

0.93 ± 0.06

4.3. Fuel burn-up

The calculation of fuel burn-up is carried out for the period from energy start-up(6 February 1984) to September 1992. In Table 3 calculation results are presentedfor the effective multiplication coefficient, the running time at 500 kW power, theaverage fuel burn-up, the maximum fuel burn-up and the U mass burn-upcorresponding to the given experimental parameters of the reactor [9]: the number

183

of fuel elements, the running time and the depth of the control rods. Table 3 showsthat the difference between calculation and experiment on Uis only about 1.9%.

235 mass fuel burn-up

Fig.4 shows the fuel burn-up values calculated for all the fuel elements inSeptember 1992. From Fig.4 it is evident that the fuel elements surrounding theneutron trap have the largest values of fuel burn-up, from 10.93% to 12.52%. Theelements near the safety rods have a smaller fuel burn-up and the lowest fuel burn-upvalues belong to the elements near the shim rods and at the periphery of the core.The minimum fuel burn-up is 5.97%.

Table 3. Calculation results of fuel burn-up.

Opera lion lime

06/02/84 - 10/8/84 a

10/8/84 - 15/1 1/84 b

15/11/84-09/3/85 a

09/3/85 - c

-

-31/12/85

31/12/85 -21/02/86

21/02/86-

-

-31/12/86

01/01/87-

-31/12/87

01/01/88-

-

-3I/12/8.X

01/01/89-6/89

6/89-31/12/89

01/01/90-

-

-31/12/90

01/01/91 -

-

-31/12/91

0l/01A>2-6/92

6/92 - 9/92

Total

Runninglime at500k W(hours)

360.5

323.2

470.2

400.0

400.0

461.8

101.5

350.0

350.0

487.3

400.0

460.0

400.0

400.0

407.0

600.0

662.8

500.0

500.0

459.6

500.0

500.0

553.6

660.0

400.0

11109

Depth of4KC(mm)

447

470

425

410

405

405

390

385

380

370

365

360

358

354

350

340

330

325

320

315

310

305

300

290

279

KcfT

1.1)026

1.0042

1.0013

1.0038

1.0030

1.0015

1.0(X)3

1.0009

1.0007

1.0015

1.0011

1.0010

1.0003

1.0000

1.0000

1.0010

1.0017

1.0013

1.0013

1.0013

1.0014

1.0014

1.0015

1.0026

1.0030

Averageburn-up

0.27

0.48

0.84

1.13

1.42

1.76

1.83

2.09

2.34

2.70

2.99

3.33

3.62

3.91

4.21

4.65

5.13

5.49

5.86

6.19

6.56

6.92

7.33

7.81

8.02

8.02

Maximumburn-up

0.41

0.76

1.29

1.74

2.18

2.70

2.82

-

-

-

-

-

-

-

-

7.06

-

-

-

-

-

-

-

11.76

12.52

12.52

Spent U235 mass (2)

Calcula-tion [9]

9.435

8.460

12.300

10.471

10.464

12.077

2.658

9.155

9.152

12.736

10.456

12.043

10.451

10.451

10.647

15.659

17.289

13.049

13.044

11.991

13.040

13.038

14.430

17.190

10.456

290.14

Experi-ment [9|

9.238

8.280

12.049

10.250

10.250

11.834

2.601

8.969

8.969

12.487

10.250

11.811

10.250

10.250

10.429

15.375

16.984

12.812

12.812

11.777

12.812

12.812

14.186

16.913

10.250

284.67

Note: a : 88 fuel elements, b : 92 fuel elements;c : 89 fuel elements from 9th of March 1985

184

Fig. 4. Calculated -fuel burn-up distributionin September, 1992.

185

4.4. Fuel reloading versions

The calculation of fuel reloading versions is based on the last core configurationin September 1992 of which the fuel burn-up distribution is shown in Fig.4. Thisconfiguration has the following parameters : the depth of the shim rods of 276mm,the excess reactivity of 3.97% (4.9 beff), the average fuel burn-up of 8.28%, thethermal neutron flux at the neutron trap of 2.29x10 * n/cm /s and the unequalcoefficient of 1.65.

10 fuel reloading versions are suggested and calculated (Table 4). From Table 4it is obvious that :

i. Thermal neutron flux at the neutron trap depends mainly on the number of fuel13 2 ' I t 2 13 2

elements and equals 2.2x10 * n/cm /s, 2.1x10 * n/cm /s and 2x10 * n/cm /s for thecore loaded with 96, 100 and 104 fuel elements respectively. So the neutron fluxdecreases with the number of fuel elements.

ii. Unequal coefficient depends mainly.on the fuel burn-up levels of the fuelelements placed around the neutron trap.

Table 4. 10 calculation versions of fuel reloading.

Para-meter

I

II

IIIIVV

VI

vnVIII

IX

X

XI

XII

XIII

Version

1

96

65

31

5.05

9.10

2.16

1.82

427

8.66

KM)

282

12.0

16.6

2

100

65

35

4.85

9.10

2.11

1.84

486

9.27

12.5

283

13.2

18.2

3

96

77

19

6.22

10.1

2.17

1.84

403

8.20

8.5

277

12.1

17.1

4

100

77

23

5.97

10.1

2.11

1.86

443

8.81

11.0

278

13.3

19.0

5

104

77

27

5.74

10.1

2.05

1.88

494

9.52

13.5

111

14.4

20.6

6

96

77

19

6.22

10.1

2.16

1.85

379

7.88

7.0

279

11.1

17.3

7

96

89

7

7.67

12.5

2.17

1.69

365

7.54

6.0

280

11.8

15.6

8

96

89

7

• 7.67

12.5

2.16

1.73

348

7.33

5.0

278

11.1

17.6

9

100

89

11

7.36

12.5

2.11

1.72

388

7.98

7.5

279

12.4

20.0

10

104

89

15

7.08

12.5

2.04

1.73

428

8.72

10.0

111

13.5

22.0

Note: I : Number of fuel elementsII : Number of old fuel elements

III : Number of new fuel elementsIV : Initial average fuel burn-up, %V : Initial maximum fuel burn-up, %

VI : Thermal neutron flux at neutron trap, n/cm /sVII : Unequal coefficientVIII: Initial depth of shim rods, mmIX : Excess reactivity, /icff

186

X : Burn-up time, 1000 hoursXI : Final depth of shim rods, mm

XII : Final average fuel burn-up, %XIII: Final maximum fuel burn-up, %

From these 10 versions of fuel reloading it is interesting to choose an optimalversion. The main criteria for choosing the optimal version are as follows :

a. The thermal neutron flux at the neutron trap is as high as possible, about2.1 xlO13 n/cm2fs.

b. The unequal coefficient is as low as possible, about 1.7.

c. The fuel burn-up of the withdrawn fuel element is as high as possible, about30%.

d. The fuel burn-up lime of the reactor is as long as possible, about 3 - 5 years.

e. The excess reactivity is about 7 faff-

On the basis of these criteria the two 7l and 81 versions have been chosen. The

parameters of the reactor core for the 81 version, in which 7 new fuel elements areinserted at the periphery of the core after having withdrawn 7 beryllium rods, areshown in Table 5 together with the parameters of the present core loaded with 89fuel elements. From Table 5 it is obvious that the new core has the maximum thermalneutron flux of less than 5%, and the excess reactivity of less than 20% in com-parision with the corresponding parameters of the present core. The new arrange-ment has a burn-up time of about 5000 hours, i.e. the next fuel reloading will becarried out in 3 years.

Table 5. Parameters of the new core after fuel reloading (8 version) and thepresent core.

Paramclcr

Total number of fuel elements

Number of old fuel elements

Number of new fuel elements

Initial average burn-up (%)

Initial maximum burn-up (%)

Maximum thermal neutron flux (n/cm2/s)

Unequal coefficient

Initial depth of shim rods (mm)

Excess reactivity (/3cff)

Burn-up time (hours)

Final depth of shim rods (mm)

Final average burn-up (%)

Final maximum burn-up (%)

New core

96

89

7

7.67

12.52

2.16xin13

1.73

348

7.33

5000

278

11.13

17.62

Present core

89

0

89

0

0

2.28x1013

1.74

490

9.57

11109

276

8.28

12.52

187

REFERENCES

1. Technical design for reconstruction and expansion of the Dalat research reactor,Vol.3, 1979 (in Russian).

2. I.E. Isakas and V.A. Shustov, HEXA-1 programme for reactor calculation withtwo- dimensional diffusion approximation, LIYAF preprint, No 490, 1979 (in Rus-sian).

3.H. Siewers, A FORTRAN program for solving depletion chain equations byanalytical methods, Atomkernergie , Vol.31, 238, 1978.

4. Yu.V. Petrov, A.N. Erykalov, Nguyen Phuoc Lan, G.R. Dik, L.M. Katova,Calculation of critical masses of the Dalat reactor in its physics start-up, LIYAFReport, No 87 ER, 1984 (in Russian).

5. Yu.V. Petrov, Resonance absorption in close arranged small blocks,, AtomicEnergy, Vol.2, 357, 1957 (in Russian).

6. E.A. Garusov, Yu.V. Petrov, Few-group calculation model of slowing-down forthe water-aluminum cores, Atomic Energy, Vol.32, 225, 1972 (in Russian).

7. Pham Duy Hien, Ngo Quang Huy, Vu Hai Long,Tran Khanh Mai, Report onphysics start-up of the Dalat nuclear research reactor, Preprint, Dalat NRI, V031-89,1989.

8. Ha Van Thong. Ph.D. Thesis, Dalat, 1990.

9. Data from Dalat reactor logbooks.

188

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993. VN9700013

KINETIC CHARACTERISTICSOF THE DALAT NUCLEAR RESEARCH REACTOR

Tran Khac An, Nguyen Nhi Dien, Pham Duy Hien, Ngo Quang Huy,Ngo Phu Khang, Pham Van Lam, Vu Hai Long, Huynh Dong Phuong,

Ha Van Tliong, Luong Ba Vien, Le Vinh Vinh

Nuclear Research Institute, Dalat, Vietnam

ABSTRACT

Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated.Experimental parameters measured consist of: Temperature coefficient of reactivity forwater-moderator. Xenon poisoning , contribution of delayed photoneutrons induced byBe(y,n) reactions and positive reactivity insertion behavior.

I. INTRODUCTION

The Dalat nuclear research reactor was rebuilt from the TRIGA Mark II reactor.The fundamental change in the reconstruction is the replacement of U-Al alloy type fuelinstead of the well known TRIGA HZr-U fuel with a high negative temperaturecoefficient. Furthermore, in order to increase neutron density, an additional berylliumreflector was introduced beside the old graphite reflector. By (y,n) reaction this materialconstitutes an internal photoneutron source that may have influence on the reactorbehavior and operation. For the above reasons, it is worthwhile investigating kinetic anddynamic characteristics of the rebuilt reactor. In this paper are presented experimentalresults concerning some kinetic parameters of the reactor.

II. REACTIVITY COEFFICIENT

For light water moderated reactor, temperature coefficient of reactivity is defined asthe ratio

aj - dp / dT ,

where dp is reactivity variation corresponding to water temperature variation dT in thecore.

For the Dalat reactor, a j has been measured at 0.25 kW in the temperature rangeof ( 20 -MO ) °C by compensation method using automatic regulating rod. The result is asfollows:

189

-(1.0 ± 0.1) 10-2|3eff/°C for T = (20-30)<>C

-(1.5 ± 0.1) 10-2Peff/°C for T = (30 4- 40)°C

With the effective delayed neutron fraction of (3eff = 0.0081 [1] we have :

= -(0.8 ± 0.1) 10-4 /°Cfor T = (20 -J- 30)°C

= -(1.2 ± 0.1) 10"4 /oC for T - (30 + 40)°C

This experimental value is not in contradiction with the design value calculated in[1], ctTca' ~ -1 10"4 / °C, and falls within the range of reactivity temperature coefficientfor research reactors moderated and cooled by water [2].

Although a y is relatively large for water moderator, its temperature effect doestake place with some delay, whereas reactivity coefficient due to variation of temperatureinside fuel elements gives rise to a more instantaneous effect. For WR-M2 type fuel ofthe Dalat reactor, this value is -0.02 lO"4/0C [1], and for TRIGA type fuel, it is - 1.3 10"4

/°C, [3]. Therefore TRIGA reactors possess intrinsic stability much greater than thepresent reactor.

III. XENON POISONING

Xenon poisoning is a negative reactivity effect due to X e ^ 5 , a (j235 fissjOnproduct having very strong thermal neutron absorption. At 500 kW nominal power,corresponding to average neutron flux of 3.4 1012 n/cm^/sec, Xenon poisoning effect inthe Dalat reactor has been measured by control rod compensation method. In Fig. 1 isshown variation of Xenon reactivity versus working time at 500 kW compared withtheoretical calculation [4].

It is seen that Xenon reactivity is negative with absolute value increasing with timeand attaining equilibrium value

p X e = - ( 1.61 ± O.O6)(3efT= - 1.3%

after about 60 hours at 500 kW.

From calculation it is also found that maximum reactivity worth after reactor shut-down, the so called Iodine pit effect, is :

= - 0 . 0 5 % ,

and the total maximum Xenon poisoning of the Dalat reactor at nominal power is -1.35 %.

190

0

-0.2

-0.4

-0.6

-0.8

-1

a -1.2

-1.4

-1.6

-1.8

Hi.CO.

i

Jrf V W V* ^ M M n. X X X X K X >•—

0 20 40 60t (h )

80 100

ig I. Varialion oficaciivily due to Xenon poisoning versus working time ofi end or at 500 kW, corresponding to neutron flux of 3 4 1 0 ^ n/cm^/sec.

o . I'.xpcrimcnt: Compiitalion

10 50 t(nun)

Time npectrum of the dclnyed neutronsafter reactor shut down.

I7.x[>cr imcuta 1 <)n la.Fitting curve for 6 fins ion neutroncomrxincnt.s nnrl 10 photoncutron componentCalculation curve Tor 6 fission neutron

components.

191

IV. PHOTONEUTRON CONTRIBUTION

In the Dalat reactor, delayed neutrons come from two contributions : neutrondisintegration of U^35 thermal fission products and neutron emission of beryllium materialinside reactor core by capture of gamma rays having energy higher than 1.67 MeV. Theproblem of determining the latter contribution, i.e. delayed photoneutron contribution, isneeded in studying reactor dynamics and in assessing neutron source remained after shut-down.

Total effective delayed neutron fraction, Pefj; consists then of two partseffective delayed fission neutron fraction from \fl^, a nd pBe^ effective delayedphotoneutron fraction from Be(y,n) reactions. Experimental data of delayed neutronspectra have been time analyzed, see [5], into 15 components, of which 6 components arefrom U^35 fission and 9 others from (y,n) reactions on beryllium. It resulted that in mediacomposed of only beryllium, delayed photoneutron fraction is 15.175 10*5, i.e. withknown delayed neutron fraction from U^35 fission pU ~ 0.0064, one has :

P B e = 2.31 % and P u = 97.69 %

In the experiment on rapid shut-down realized on a critical assembly with cylindricalcore composed of enriched uranium ( 10 % Tj235 ) an(j beryllium reflector; Be/Tj235 =1777 -=-3112, [6], one obtained delayed photoneutron fraction of 14.808 10"5, so

(3Be = 2.26% and P u = 99.74%

In the Dalat reactor, time spectra of delayed neutron were measured after reactorshut-down following a continuous run of 100 hours at 500 kW. Neutron spectra,measured by fission detector KNK-15 of the reactor control system AKNP and by IN-90acquisition computer, are compared with indium foil activation data. In Fig. 2 timespectrum of delayed neutrons during 64 hours after reactor shut-down, normalized to unitat the moment of shut-down, presents sharp variation at the beginning. This rapid part ismainly contributed by delayed neutrons from Tj235 fission products, after about 10minutes this contribution can be neglected in comparison with delayed photoneutrons,slowly varying neutron source of intensity about IO~4 4. 10"? times the neutron sourcelevel at the moment of shut-down. So the photoneutron source left after reactor shut-down gives the neutron flux of about ( 1 0 ^ - 10^ ) n/cm^/sec. This extraneous neutronsource is used for starting-up the reactor.

A global estimation from time spectrum gives

P B e = 0.46 % and P u - 99.54 %

These data show that p B e / P u ratio is smallest for the Dalat reactor ( 0.46 % ),whereas it is the biggest for reactor moderated only by beryllium (2.31 % ). This can beexplained by the fact that beryllium reflector of the Dalat reactor is only a supplementary

192

500 .

5 0 . 0

5 . 0 0

0 . 5 0

0 . 0 5

p = + 0.38 Pefr

- \ / ^ -

/ y/ / P

V-.„ ! . . „ 1, 1....

11

V

- 1- 0.27

i

. — •

PeflT

. i . i i i i

1 2 S 9t ( mill. )

- Dependence of reactor power on time after inseitionof positive reactivities.

600

500

400

300

200

100

0

s

yY7 \

< /

.J ' v ...,,...—,.

> }

0 0.1 0.2 0.3

p(ncfr)0.4 0.5

I'ig. A. Dependence ofstablc power levels attainedon positive reactivily inserted.

193

reflector composed of some Be-plates and Be-rods disposed separately on thecircumference of the graphite reflector ( weight ratio Be/U^-^ =: 8 7 )

V. POSITIVE REACTIVITY INSERTION BEHAVIOR

A nuclear reactor is a controlled system with feedback. From the dynamics point ofview, it is then important to study stability property of the reactor under modification ofdynamic parameters.

For its behavior studied during the short time after a rapid insertion of positivereactivity, Xenon poisoning effect can be neglected, reactor stability is then assured onlyby negative feedback of moderator and temperature effects.

For the Dalat reactor, stability has been studied at 0.1 kW for different values ofreactivity inserted by withdrawing control rod. In Fig. 3 is shown the increase of reactorpower versus time for inserted reactivity of +0.27 (3eff and +0.38 3eff anc* t n e p o w e r

levels attained after 10 minutes due to negative feedback of water and fuel temperatureeffects. Fn Fig. 4 is shown the dependence of attained stable power levels on insertedpositive reactivity.

From this study it is confirmed the safe operation specification, i.e. in order to avoidreactor shut-down by over-power condition it is necessary to limit positive reactivityinserted to value of less than 0.4 (3eff.

Furthermore, as already mentioned in Section IT, TRIGA reactor has a stabilitybehavior better than the Dalat reactor because of TRIGA's higher negative instantaneoustemperature coefficient. Indeed, TRIGA reactor can support inserted positive reactivity of1 (3eff resulting in reactor period of 1 sec and reaching only 120 kW stable power levelafter having attained several hundred MW [4].

VT. CONCLUSION

The main results of the investigation of the kinetic characteristics of the Dalatreactor are as follows:

1. The temperature coefficient of reactivity for water-moderator is about - ( 0.8 -rl .2 ) 10"4 / °C for temperature range of ( 20 4- 40 ) °C.

2. The reactivity coefficient of Xenon poisoning effect is -1.3 % .

3. The effective delayed photoneutron fraction from Be(y,n) reactions is 0.46 %.

4. The safe positive reactivity inserted is of less than 0.4 (3ejgf.

194

5. The Dalat reactor possesses an intrinsic stability much less than the TRIGAreactor because the instantaneous temperature coefficient of reactivity and safe positivereactivity inserted of the Dalat reactor ( -0.02 10~4 / °C and 0.4 (3eff) are much less thanthose of the TRIGA reactor ( -1.3 10"4 / °C and 1 (3eff).

VII. REFERENCES

1. Technical design for reconstruction and expansion of the Dalat research reactor,Vol. 3, 1979 (in Russian).

2. Directory of Nuclear Reactors, Vol. 2, IAEA, Vienna, 1959.Directory of Nuclear Reactors, Vol. 3, IAEA, Vienna, 1960.

3. Technical Foundation of TRIGA, General Atomic Co. , GA-471, 1958.4. I. X. Ganev.

Physics and calculation of reactors, Energoizdat, Moscow, 1965 (in Russian ).5. G. R. Keepin.

Physics of Nuclear Kinetics, Addison Wesley, 1965.6. A. K. Krasin at al.

Proceedings of 2nd UN Conference on PUAE, Vol. 2, p. 571, Geneva 1958.

195

mmmmmmmmmi

•HEXT

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993. " ™ " S o b ™ '

EXPERIMENTAL DETERMINATION OF FUEL SURFACE

TEMPERATURES IN THE DALAT NUCLEAR RESEARCH REACTOR

+ Nqo Phu Khang, Nqo Quang Huy, Tran Khac An,Pham Van Lam, (Nuclear Research Institute,Dalat - Vietnam)

+ P.A. Diakov, A.F. Sacsanov, (The USSR StateCommittee on the Utilization of Atomic Energy)

ABSTRACT

Measured fuel surface temperatures,obtained at variouslocations of the core of the Dalat nuclear research reactorunder normal operating conditions, are presented, and somethermal characteristics of the reactor are discussed.

1. INTRODUCTION

The Dalat nuclear research reactor is a 500 kW pool reactorwith Soviet WR-M2 fuel bundles cooled by natural convection. Thereactor core, designed in the Soviet Union,is installed in a 2.0meter diameter,6.2 meter deep aluminum tank.

The core consists of 89 fuel bundles, a central neutron trap,and three experimental channels. Each fuel bundle is composed ofthree concentric layers (see Fig.l): 2 cylindrical inner tubes anda hexagonal outer tube. The cladding material is 0.9 mm thickaluminum. The weight of U-235 in each fuel bundle is approximately40 gram. . '

On the thermal hydraulic point of view,500 kW is a relativelyhigh power for natural convection cooling. However, cooling wasenhanced by the addition of an extracting well in 1982 during thereconstruction (see Fig.2).Water boiling at a fuel element surfaceis harmful and can lead to a dangerous situation. Consequently,thesurface temperature of a fuel element must be maintained wellbelow the temperature at which surface boiling occurs. Therefore,careful measurement of fuel surface temperature can assist withboth nuclear safety and prospective power upgrading studies.

In this paper, an apparatus for measuring fuel surface tempe-rature is described. Some vertical and radial fuel surface tempe-rature distributions,obtained under various reactor conditions(power level,relative positions of the shim rods,etc.)/arepresented. Experimental and calculated fuel surface temperature atone of the hottest positions of the core are compared. Finally,some general assessments of the thermal characteristics of thereactor are presented.

197

Flq.J. Fuel bundle typ<>

Flq.2. Schematic dlAarom ol the Oalat ttactot coollnr) system

1- Reactor cora, 3- Extracting wall, )- Support plata ofexttactlna wall, 4- (traphltt reflector, 5- Reactor tank,<- Position of Inlet primary loop pipe In the tank, 7- PrimaryDumo. I- Mpat otrhanotr, 1- Secondary sum. 10- Conllnq towt>r.

Flq,3. Instrumented (u»l bundle with, thermocouples for mn4aur*i*entof fuel surface temperature

riq.4. Incorporation schema of thermocouple In alumlnimcladding of Instrumented fuel bundle

0.8Cables of fchormocoupls

0.9

0.7

Isolator

Junction

Stainleao steal oovsr

198

2. EXPERIMENTAL INSTRUMENT

The device used to measure the fuel surface temperaturesconsisted of an instrumented fuel bundle (IFB) together with anelectronic measuring system(see Figs.3-5). The IFB is of the sameWR-M2 type as normally used in the Dalat reactor. Nine Chromel-Coppel thermocouples were incorporated into the IFB. Of these,fivewere incorporated in the aluminum cladding of the hexagonal outertube; two were incorporated in the aluminum cladding of the twocylindrical inner tubes and the remaining two were in directcontact with water to measure the core coolant inlet and outlettemperatures.

The location of the 9 thermocouple junctions along the IFB isas follows:

Junct ionnotation

JiJ2J3J4J5J6J7

J8

J9

Temperaturenotation

TiT2T3T4T5T6T7

T8

T9

Distance fromIFB centreline

(mm)

230

180

130

0

-150

210

210

345

-355

Remark

-Hexagonal outer tube•f

tl

it

ti

-Middle cylindrical tube

-Innermost cylindrical

tube

-Measurement of the core

outlet water temperature

-Measurement of the core

inlet water temperature

The junction of each thermocouple (other than J and Jq,whichmeasure the water temperature) was buried 0-3 mm inside the 0.9 mmthick aluminum cladding of the IFB. The presence of thermocouplesin the IFB did not affect its thermal hydraulic performance

measures durino construction fl).thanks to careful

The supporting electronic system consisted of an analogprocessing unit ( amplifier plus analogue-to-digital converter )interfaced with an AT personal computer. A block schematic of thesystem is shown in Fig.5.

199

Pl i j .5 . n I tick ncln-mw ol neAaiirln't system

——

1/1

t/J

1/J

—

FI«l.S. l.»y(inl: n( tin? Oitl.it rei icloi coti.

J rual l>un.ll«

Cuntrol toil

Btr y I I I IM block

11 r.xllxt Ion rbnn«r\ e Poal'. Ion u.t«IM* A 9 • i r n m^ n t *

200

3. RESULTS

Temperatures were measured at selected locations in thereactor core. At each of these locations,the normal fuel elementwas replaced by the IFB and thermocouple temperatures monitored.Ateach location, 6 different measurement sets could be obtained byrevolving the hexagonal IFB around its axis by angles of 60 .Afterthe IFB was placed in the core,the reactor was operated at a fixedpower level. Temperatures T. - T_ were monitored and recorded bycomputer. Fig.6 shows the layout of the reactor core, and thepositions where measurements were made.

3.1. Axial fuel surfacetemperature distribution

This distribution was determined for the outer hexagonal tubeof the fuel element at cells 9-6, 6-5 and 5-1 (see Fig.6). Themeasured temperatures T. - T- along the core height are shown inFigs.7, 8 and 9 for various reactor power levels between 50 and500 kW.All curves have th-e same shape with the maximum temperatureposition near the center of the core. •

The effect of shim rods is shown in figure 7. Both thetemperature distribution and maximum temperatures vary slightlywith shim rod position.

3.2. Radial fuel surface temperature distribution

The radial distribution of surface temperatures over thereactor core was obtained from measurements carried out at cells5-1, 6-3, 6-5, 5-6 and 9-6. Fig.10 shows the T. data (i.e.maximumtemperature at each fuel element location) for a reactor power of500 kW. It can be seen that the highest T. temperature is reachedat one of the hottest positions in the core (cell 5-6)close to theneutron trap.

The radial temperature distribution within a fuel element wasobtained from data from thermocouples T_, T_ and T_ which arelocated on the three separate fueled sections of the IFB. Thisdata,as obtained at cell 5-6,is shown in figure 11. This figureindicates that the highest temperature is on the outer surfacenear the neutron trap.

The data obtained at different locations in the core indicatethat cells 5-6 and 9-6 are at the hottest positions, and thesurface temperature at these positions reaches a maximum of about99 C on the hexagonal outer tube for the reactor operating atnomi na1 power.

3.3. Variation of maximum waterand fuelsurface temperatures with

reactor power 1 eve1s

Fig.12 shows the dependence on power level of fuel surface

201

l l rrl\ f-fi for »nr(oii« f r«>«olor p»««r

• 4 -tlllm rnrin nt Mm nnra* hnlr.ht

• 4 "him m i l l nt •lirfxrxn* h«(r.*<t

- )f>O -?0O -I0O 300

Tf""P«T>llllr»(Or)

P-500 mr

P.50

-?oo -100 100 inn joo in-™)

202

T-500 K«f

- O - -

r-)oo r t

-—©

— o

_joo -?oo -loo o ioo ?oo joo

Ttmperntur

i t rpnctor pn»nr of 500 kW

trnp

1O0 ?00 JOO 400 500 f(Vt)

203

temperature T and water temperature T_ at cells 5-6 and 6-3.It is clear that these temperatures increase proportionally withthe reactor power at lower power levels and change more slowly atlarger power levels.

4. COMPARISON OF PREDICTED AND MEASURED TEMPERATURES

The method of predicting thermal hydraulic characteristics ofthe Dalat reactor can be found in [2]. The following table givespredicted and experimental maximum fuel surface temperature forvarious power levels.

Reactor power(kW)

50

100

200

300

400

500

Maximum fuel surface temperature! C)

Theoretica1

39 .56

48.34

62.62

74.76

86.37

98.71

Experimental

41.0

4 7.0'

62.0

74.0

8 6.5

99.0

While IFB measurements could only made at some restrictedlocations, theoretical calculation can give temperature valuesat any position in the reactor core height. The good agreement ofmeasured and predicted temperatures at locations where the IFBcould be placed adds confidence to the validity of predictedtemperature at any location in the core. Since the IFB was placedat locations of particular interest,the agreement of predicted andmeasured temperatures here indicates that we can be particularlyconfident that the prediction method is reliable at theselocat ions .

5. CONCLUDING DISCUSSION

Experimental temperature measurements show that maximum fuelsurface temperature at the hottest cell of the Dalat reactor canreach nearly 100°C when the primary water temperature at the

204

reactor tank inlet has attained its highest permitted value of30°C. Although 100°C is slightly larger than the upper limit, itis far below temperature T at which surface boiling occurs.However, 100 °C should not be exceeded since during the reactorstart-up, measured fuel surface temperatures reached a peakapproximately 2 °C higher than the steady state peak. According tothe designer's documents,the scram limit is set at 110% of nominalpower. In this case,the fuel surface temperature would reach avalue of about 108°C, which coincides with the saturated watertemperature in the reactor core (but is still welj below thetemperature T at which surface boiling would occur T =116°C (21).s s

The experiments also indicate that the temperature of waterand fuel surface quickly become stable during start-up. This is adesirable feature of the Dalat reactor.

REFERENCES

[I). Technical documents of the experimental apparatus for fuel

surface temperature measurement of the Dalat nuclear

research reactor, The USSR state committee on the

utilization of atomic eneray,Moscow 1989.

[21. Nqo Phu Khanq : Study of thermal hydraulic characteristics

and possibility of power upqradinq of the Dalat nuclear

research reactor,Int.Rep., Dalat 1987.

205

Proceedings of the 4th National Conference on PhysicsHanoi, Vietnam, 5 - 8 October 1993.

VN9700015

THERMOHYDRAULIC CHARACTERISTICS UNDER SOME TRANSIENT

CONDITIONS OF THE DALAT NUCLEAR RESEARCH REACTOR

Quang Huy,Ngo Phu KhangTran Khac An,Huynh Ton Hghlem

ABSTRACT

Some experimental and theoretical thermal hydrauliccharacteristics of the Dalat nuclear research reactorare presented, together with *ome general assessments,from a chermal hydraulic point of view, of its safetyunder transient conditions.

1. INTRODUCTION

The experimental device used for measuring in-coretemperatures of the Dalat nuclear research reactor has beenoperatinq since 1990 to achieve its objectives, namely, theexperimental determination of water and fuel surface temperatureunder normal and transient conditions of the reactor,required aspart of thermal hydraulic study of some nuclear safety aspects ofthe reactor view. Some recent experimental data were obtained fromthis device 11), allowinq some significant conclusions about thesafe operation of the reactor under normal workinq condition to bedrawn.

Further assessment of the Dalat reactor safety requires astudy o-f transient processes. Nearly all reactor dynamic behavioris closely coupled to the water and fuel surface temperaturedistributions in the core. The technical design documentation [2],and theoretical thermal hydraulic calculations performed afterreactor inauguration f31,give some results oE various hypotheticalaccidents (reactivity accident,loss of coolant flow accident,etc ).However,using the instrumented fuel bundle (IFB) into which 9Chromel-Coppel thermocouDles have been inserted,we can obtaintime-dependent: temperature data not only for rapid insertion of asignificant positive reactivity in the core but also duringtransient processes such as chanqes of reactor power,water flowrate in the primary loop,and temperature at reactor tank inlet. Inthe following, some experimental thermal hydraulic characteristicsare presented for the Dalat reactor under transient conditions.The paper outlines a method used to calculate transients whichoccur if the reactor runs without its cooling system. Finally,experimental and predicted transient temperature data arecompared,and some assessments are made of the Dalat reactor'ssat'ity during transient conditions.

207

2. EXPERIMENTAL RESULTS

A description of the instrumented fuel bundle together withan electronic measuring system can be found in fl). The transientcases considered in this paper are :

- Manual step by step reactor start-up- Automatic reactor start-up- Planned reactor shut-down- Reactor scram due to primary pump failure- Reactor operation at 300 kW power with no cooling system- Reactor operation at 200 kW power with no cooling system- Reactor operation at 200 kW power following the reinfcro-

duction of the cooling system after its non-operationfor a few hours.

Figures 2 - 8 show experimental temperatures and power as afunction of time at cell 9-6 [see Fig.l] of the core for all theabove cases. The notation of these curves is as follows:

i ( i=l,2,...,7 ) : fuel surface temperature T. ( C)i ( i = 8,9 ) : water outlet and inlet temperature ( C)P : reactor power ( % of nominal value )t : t i me

Figs.2 and 3 show that during reactor start-up,water and fuelsurface temperatures quickly become stable. Therefore, asdescribed in the reactor operating instructions [41, automaticinstead of manual start-up is completely acceptable.Figs.4 and 5show rapid decreases in power and temperatures,except for thetemperature at the inl°.t of the IFB, after reactor shut-down.Experimental results obtained when the reactor ran without itscooling system at 200 and 300 kW are shown in Figs.6 and 7.In thiscase, temperatures T. (i=1,2, . . . , 9 ) increase almost linearly withtime. Fig.8 presents data obtained when positive reactivitywas rapidly inserted in the core when operating at 200 kW. Thisevent was achieved by turning on the cooling system,i.e. insertingcool water in the core,following a period when the cooling systemwas not used. This abnormal situation has a significant effect onthe safety of the reactor and it must be considered seriouslyduring operation.

3. CALCULATION RESULTS

3.1. Model

In order to obtain an estimate of the maximum water and fuelelement surface temperatures for the case of the reactor runningwithout its cooling system, a simple thermal hydraulic model wasdeveloped. The model considers reactor core cooling by naturalconvection,and makes the following assumptions :

- Hydraulic processes in the reactor tank are guasi-stationary.

208

- Heat transfer from the tank into the surrounding air is mainlythrough the open surface of the reactor tank water.

- The temperatureconstant ( T.

K K

of= 20°C )

the surrounding air T.. is assumed to bekk

According to this model,the water flow rate through the reactorcore is given by : *

9.81 H = P (1)

With P l = (X1H1

P2= (X2H2

P4=

P5=

(G/S2)2/2r2

(G/S5)2/2r5

In the above

piG

H.

D.

S.

X.1

hydraulic loss in space i

water flow rate throuah the core

reactor core height

extracting well height

water density at temperature T.

equivalent hydraulic diameter of space i

flow area throuqh space i

friction coefficient of space, i

local hydraulic resistance of space i

(2)

(3)

(4 )

(5)

The heat balance in the reactor tank can be written as:

a F f T - T )b b{ b kk;

Where

c (T. ) m.dT./dtp b b b = P (6)

kk

m.

heat transfer coefficient through open surface of the

reactor tank water

area of open surface of the reactor tank water

average temperature of the reactor tank wster

temperature of the surroundinq air

water specific heat at constant pressure

mass of water in the reactor tank

reactor power

209

F l a . l . Layout ot l.h«

l^ J m» I bund I *

Control rort Noutron trap

e pos it Inn HIMwisytnanta

j.J. mninl Urn by it«p mctot »tact-ap

210

Agto—tlc

4«« t lavcl

a t i»«ci

211

. j . M«ctot »ci»» •»»« to ptiMt

21 tl»cl

J06 >w power with

cool ln<» » y » f

212

100 «w p o v r

no cooling »wt«•

rig.I. »«»ctor op«r»tlon »t tBO IrW p e w t tollowlnn

ch« g>iwtco**uctlon af th» cooling l y t w

*lt«r lt» non-op«r«tlow tot * C«v hour*

•o ?nn ion ««n (i«"i

213

_ K

0T

tyrttim

n*. 7- Prl«cy

214

The axial coolant and surface temperatures for the hottestfuel element at any moment t can be given by :

- Water temperature

- Fuel surface temperature

dT(z)/dz « 4q (z)/c5

T (z)5

Where qs(z) surface heat flux

ma s s rate

heat transfer coefficient

= T(z) q (z)/hs

(7)

(8)

3.2. Results

Tables 1 - 5 give the predictions for the reactor running atvarious power levels without its cooling system. The tank watertemperature at the beginning of each run is assumed to be 25 C.Tables 6 and 7 compare the predicted and experimental temperaturesfor those data illustrated on Fiqs.6 and 7.

Table 1. Predicted temperatures for reactor operation at 500 kWwithout cooling system

Time(min.)

01020304050

Maximum temperature ofwater in the core f C)

50.552.955.558. 261.164.1

Maximum fuel surfacetemperature [ C]

96.798.6

100.7103.1105.6107.7

Table 2. Predicted temperatures for reactor operation at 400 kWwithout cooling system

T i me ( m i n . )

020406080

100

Maximum temperature ofwater in the core (°C1

47.951.956. 461.266 . 371.9

Maximum fuel surfacetemperature [ CJ

86.389.793.697.9

101.9105. 4

215

Table 3. Predicted temperatures for reactor operation at 300 kWwithout cooling system

Time(min.)

04590

135180

Max imumwater in

temperature othe core (°C

44.952.260.669.679.5

f1

Maximum fuel surfacetemperature ( Cl

75.181,689 .197.1

104.4

Table 4. Predicted temperatures for reactor operation at 200 kWwithout coolina system

Time(min.)

060

120180240300

Maximum temperature ofwater in the core I C 1

41. 348.155.763.772.181.1

Maximum fuel surfacetemperature [°C1

62.668.975.983.5

- 91.599 .2

Table 5. Predicted temperatures for reactor operation at 100 kWwithout coolinq system

Time(min.)

p

120240360480600

Maximum temperature ofwater in the core ( Cl

36.743.951.960.268.877.6

Maximum fuel surfacetemperature [ C1

48.455.262.871.079. 387.9

216

Table 6. Comparison of theoretical and experimental temperaturesfor the case of the reactor running at 300 kW withoutcoolina system

Time(min.)

7142128354249566 3707784

Maximum temperature ofwater in the core I C]

Theoretica1

46.147.248.349.550.751.953.154. 455.756.958.259.6

Exper imental

50.050.852.453.655.657.258.060.861.862.265.265.4

Maximum fuel surfacetemperature ( C)

Theoretical

76.177.178.079.180.181.282.38 3.584.685.887.088.2

Exper imental

75.677.178.579 .980.981.983.384.485.786.688.089.2

Table 7. Comparison of theoretical and experimental temperaturesfor the case of the reactor runninq at 200 kW withoutcoolina system

Time(min.)

102030405060708090

100110120

Maximumwa t e r in

temperature ofthe core (°C1

Theoretical

44.745.847.048. 249 . 450.651.953.154 . 455.757.050. 3

Experimental

47.348.250.052.153.954 . 255. 456.858.759.661.162. 4

Maximum fueltemperature

Theoretical

65.766.767 .868.970.071.272. 473.674 .775.977.278. A

surface(°C1

Exper imental

64.866.267.469.070.571.472.873.975.176.377.478. 5

217

4. CONCLUSIONS

Significant experimental and theoretical transient thermalhydraulic data have been obtained for the Dalat reactor. As aresult of natural convection rooling,water and fuel equilibriumtemperatures are quickly reached during reactor start-up and shut-down. In the absence of the primary coolant pump,natural convectioncooling by the tank water is sufficient by itself to assure goodcooling of the core after shut-down. Moreover,this coolingmechanism is sufficient to allow the reactor to be operatedwithout the heat exchanqe system for some considerable time. Thistime depends on reactor power,and may be several hours at lowerpower level.

REFERENCES

[11 Ngo Phu Khana et al. : Experimental determination of fuel

surface temperature of the Dalat nuclear research rractor,

Int. Rept.,Dalat,1990 .

[2). The USSR state committee on the utilization of atomic energy:

Technical design documents of the Dalat reconstructed

reactor,Moscow 1979.

f 31 . Nuclear research institute (Dalat) : Safety analysis report

of the Dalat reactor, Dalat 1989.

218

VN9700016

EXPERIMENTAL INVESTIGATION

OF THE NEUTRON PHYSICS CHARACTERISTICS

OF THE DALAT NUCLEAR RESEARCH REACTOR

Ngo Qunng IluyConter of Nuclear Techniques, Ho Chi Minli, Vietnum

Ha Van ThongNuclear Research Institute, Dalat, Vietnam

ABSTRACT

The Dalat nuclear research reactor was reconstructed from the TRIG A MARK]] reactor installed in 1963 with a nominal power of 250 kW. The now reactorreadied its planned nominal power 01*500 kW for the first time in February 1984.Investigation of the neutron physics characteristics of the Dalat reactor has beencarried out in the course of its operation. The main result: obtained is as follows:

- The effective fraction of delayed photonoutrona and the extraneous neutronsource left after reactor shut down are measured.

- The lowest power levels of critical states of the reacLor are determined.

- The perturbation effect is investigated when a water column or a plexiglassrod is substituted for a fuel element.

- Tho relative axial and radial distributions of tho thermal neutrons measuredand the geometrical parameters of the core such us the in homogeneous coefficients,the bucklings, the effective height and radius, tho extrapolated distances areobtained.

- The thermal neutron distributions are measured around the old graphitereflector.

1. INTRODUCTION

The Dalat nuclear research reactor was reconstructed from tho TRIGAMARKII reactor installed in 1963 with a nominal power of 250 kW. In 1975 all the fuelelements of the TRIGA reactor had boon dismounted and taken awa}', whicheffectively caused the reactor to stop its operation. Reconstruction of tho reactorwith a higlurr operation power of 500 kW was accomplished with n Russiantechnological assistance. The new reactor reached its planned nominal power of500 kW for tho first time in February 1984.

Tho Dalai, rend-or is unique of Its kind in tho world: Tho Russian-designedcore it; harmoniously integrated into I.ht» lofl-ovor infraiibructuro of tho

219

J

\

tl»vitii>n of rune to • .'rys

I . <:n>-cZ. f.|ii|ililtc nct'lecto •

}. Sucl:i"C Tubo

4. l'rli««Vy C<»olinc .'Ivaton

$. Additional Shi old-.nr..

-Z H««ctor 'lorn Ai-i-*n*.«m<int

l;«f<ity vol

I rrniHnt ion

i n w i l i e Ir.-tii

J ll'iolroti

220

American-made TRIG A research reactor. The undlsmounted components of theformer reactor are the reactor tank, the concrete: shielding, the graphite reflector,the beam tubes and the thermal column.

The new core is loaded with the Russian-mado VVR-M2 fuel elements ofuranium-aluminum alloy. An amount of beryllium material is placed at the centerof the core, forming a neutron trap, and at the periphery of the core, forming asupplementary reflector. The existence of beryllium material introduces anextraneous neutron source induced by Be(y,n) reactions.

The extraneous neutron source enhances power levels of critical states of thereactor in neutron physics experiments of up lo sevoral hundreds watts. It leadsto the fact that the measurement of neutron physics parameters in this reactoris more difficult than in a critical assembly or a "pure" reactor in its physicsstart-up stage, where experiments are carried out at power levels of a few watt.

• The paper presents the investigation result of Lhe neutron physicscharacteristics of the Dalat reactor: The effective fraction of the delayedphnUmeutrons, the extraneous neutron source left after the reactor is shut down,the lowest power levels of reactor critical stales, the neutron field perturbationeffect, the relative axial and radial distributions of thermal neutrons in the coreand around the graphite reflector.

2. REACTOR DESCRIPTION

A view of the Dalat reactor is presented in Kig.l, in which new componentsconsist of the reactor core, the extracting well and the cooling system. The reactorcore of a cylindrical shape with 43.(i cm diameter (Fig.2) is surrounded by theold graphite reflector of 34.0 cm thickness. The reactor Is loaded with 89 fuelelements of VVR-M2 type. A fuel element has an overall height.of 86.5 cm including60 cm uranium-aluminum fuel part (36% enriched U ' >) that consists of threecoaxial tubes of 2.5 nun thickness. A tube is made of a 0.7 mm fuel layer sandwichedin by two aluminum cladding layers of 0.9 mm thickness (Fig.3). The averagemass of U ' in each fresh fuel element is 40.3 g.

Irradiation facilities inside the reactor core consist of a neutron trap, a wetchannel and two vertical pneumatic transfer tubes. The reactor is controlled bysix boron carbide rods (two safety rods AZ1-AZ2 and four shim rods KC1-KC4)and an automatic regulating stainless steel rod All.

An additional beryllium reflector layer is inserted between the active core (thefuel region) and the graphite reflector. The equivalent radius of the active core is18.9 cm and the average thickness of the beryllium layer is 2.9 cm.

3. DELAYED PIIOTONEUTRONS

Delayed neutrons in the Dalat reactor come from two contrlbullonfl: the delayedneutrons from U ' ° fission ami the delayed |)holniu!Ul,n»niJ7()',n) reactions Inducedby gamma rays of energy greater than I.CWi MnV on beryllium material.Determination of this delayed pholoncutruiiH plays nn Important role in studying

221

cur. w»_

SEJ

Pitt. 3. : Fuel element of VVR-M2 type

Lo«(H( t)/Ho) 10 50 t(pfa)

A 1 1 ,

*(b]

• Time spectrum of the delayed neutronsafter reactor shut down.

: Experimental data.: Fitting curve for 6 fission neutroncomponents and 10 photoncutron components

: Calculation curve for 6 fission neutroncomponents.

222

the reactor kinetic property and estimating the neutron source loft after tho reactoris shut down.

The time spectrum of the delayed neutrons is measured when the reactor isshut down after 100 hours of continuous operation at 500 kW power (Fig.'l). Fig.4shows that the delayed U *d fission neutron counts decrease rapidly and theymay be neglected at 10 minutes after reactor shut down. The delayed photoneutroncounts decrease more slowly and remain about 10*' -i-10* of nominal power during60 hours. So tho photonoutron source left after reactor shut down gives tho nqutron

K Q O

flux of about 10 -r 10 n/cm /s during 60 hours.The delayed neutron spectrum is analysed into 16 components that consist of

235

6 components of the delayed U fission neutrons and 10 photonoutroncomponents. The total effective fraction of tho delayed neutrons/?eff consists thonof two parts: /ierf , effective fraction of the delnyod fission neutrons and /Jerf

e,efi'ectivo fraction of tho delayed photoneutrons. A global estimation from tho timespectrum gives:

/*elTiO = (0.49 ± 0.02)%/?cff

Table 1 presents the /3ujf o//?or|- ratio function of the mnc/mij mass ratio for

several various media. It is obvious that the /Jerr °//Wf ratio increases with thoinBo/mU m a ss ratio.

Table 1. /ieff °//Jorf ratio as function of lnjju/mu mass ratio

8.51777-3112

Pui^e Ixjryllium

0.49%2.20%2.31%

Itofuruncxj

This workU l[21

4. LOWEST CRITICAL POWER LEVELS

The determination of lowest critical power levels of tho Dalat reactor containingan extraneoiis source is interesting from the view-point of safety in the reactorstart-up process and from the requirement of finding possible low powers forneutron physics experiments.

The experiment was carried out in March 1992 at 10 days after the reactorwas shut down following 100 hours of its continuous operation at 500 kW power.The two safety rods were withdrawn completely, the four shim rods wore placedat 29.5 cm in depth and the regulating rod at the «'J5 cm. Tho reactor power was6.7 10" M*,!, whore I'u is 500 kW nominal power. Tbo reactor powor was increasedby shooting a cadmium sample from tin1, read or ton1, using i\ pneumatic transfersystem. It is equivalent to an insertion of a positive reactivity of /> — 4.67% (1^.

Fig.5 shows the neutron density response! to tho posltlvo step changes inreuctivity, Tho density CUTVOB in Biibcrlticnl fitntos incronoo and attain tho eUiblevalues. They are well described by the subcritinil reactor kinetics theory [3] with

223

5 GOO

6OOO

4000

HOOO

3SOO

3OOO •

26OO

ok-l . 04.£>

2O <*O 6O 8O 1OO 12O 1-4O 100 200

Fig. 5 . Subcritical neutron density response to th« positistep changes in reactivity CO m *

•*• : ELxperiwwntal data (March lOth.- - - : Calculation with 0 deiayerd neutron groups,

. - • - • - i Calculation with 10 delayed neutron

-20,

ICT

sc5

1 O -

<b)

'. V * 1 1 )

2 D 4 5 « 7 6 9 10 11 12 13 M IOTinta «rt«r reactor 'B<"ut uowti (day)

Fig 6. Lowt<i5t c r i t i c a l power l e v e l s as function of tiiiua a f t o rtb* r snetor i s shut down.

I : Bxper Inwnt,: Calculated valu>> at Pc so than Clw first; Kzrn on

the r ight hand of the formula (2) aquals 3©«% ( A ) . L«®%<"o) , Iv»X (c> and % vd> oi th*» second ona during

224

the positive step changes in reactivity of/5 = 4.67% /Jon».

The critical state of the reactor is observed when the four shim rods are locatedat 28.4 cm in depth and the regulating rod is withdraw continuously from thedepth of 40 cm to the depth of 20 cm. The lowest critical power depends on thebeginning depth of the regulating rod. The values obtained are from 1.6 10" P n

to 4.2 10" Pn for the beginning depth from 40 cm to 60 cm of the regulatingrod.

The range of the lowest critical powers depends ngnln on the tiino after thoreactor is shut down (Fig.6). They have the values from (0.5 + 1.2) 10" Pn, i.e.25 W + 00 W, to (1.1 + 1.6) 10"5 Pn, i.e. 5.5 W + 8 W, at 4 days to .13 days afterreactor shut down.

In the just critical situation, k=l , the neutron density or power level will riseat a linear rate [3]:

P(t) = t + Pc (1)

where P8 = (1.93 ± 0.01) 10 Pn is extraneous nourcn power, P c is critical powerwithout extraneous source, 1 is prompt neutron lifetime, ftm and Xm are delayedneutron fraction and decay constant for the i 1 precursor.

10 days after reactor shut down the reactor power can bo expressed ns follows:

P(t) = 5 10"8 P n t + Pc (2)This formula shows that during the first 1000 sec in the critical situation the

first term on the right hand is greater than 300% of the second one of 1.6* 10"°Pn.It means that the reactor power is not just constant but rapidly increases. Inorder to decrease the power variation to the values of less than 10% it is necessaryto take critical power of 5 10" Pn.

In Fig.6 the critical power is plotted against the time after the reactor is shutdown so that the first term on the right hand of formula (2) equals 300% (curvea), 100% (curve b), 10% (curve c) and 5% (curve d) of the second one during 1000sec. From Fig.6 it is obvious that the lowest critical power measured varies intho range of (100 -•• 300)% during 1000 sec. These variation limits are too largefor practical use of tho reactor. They have to be reduced to tho values of about (5.-r 10)% to sustain the stable crlUcality of the reactor. Therefore critical powermust be chosen greater than 1 10"' Pn or 500 W.

5. NEUTRON FIELD PERTURBATION EFFECT

The measurement of the distribution of neutron Hold in tho Dalat reactor isunable to be carried out by activation of foils placod on the surface of fuel elementbecause of high power levels of the reactor as nicntlonod In Section 4. It is possibleto replace the measurement of the relative distribution of neutron field on thesurface of fuel element by the measurement in a replaced malarial if one provesthat the replacement induces a perturbation of the neutron fiold.

225

The perturbation of neutron field leads to the fact that tho relative change ofboth the neutron density Noon the surface of fuel element and the neutron densityin replaced material, A = (No - N) / No, is constant, depending only on the typoof the replaced material and noil depending on the position of the fuel element[4j. For tho Dalat reactor it is necessary to check the perturbation effect of thoreplacement .of the fuel element by another material, for example, by a watercolumn or a plexiglass rod.

The experiment is carried out for the reactor core loaded with 85 fuel elements.Tho axial distribution of the thermal neutrons is measured at 5 water columnsby the activation of Cu foils at 2.5 kW powor. After that a solid plexiglass rodhaving the same outer size of a fuel element is placed in a water column. Thomeasurement of the axial distribution of thermal neutrons inside the center lineof the plexiglass rod is carried out for these 5 water columns. Experimental resultshows a similarity of each pair of relative distribution curves of thermal neutronsfor each water column and replaced plexiglass rod.

The coefficient A is constant, in the error range for these 5 measured positionsand equals about 10%. This result is of great importance because the size of fuelelement can not be considered very small in comparison with the size of the core.Furthermore, investigated water columns are distributed at different locations ofthe core: two close by the neutron trap, two close by the beryllium reflector andone at the middle of the active core.

The perturbation effect induced by the replacement of the plexiglass rod forthe water column is verified by two independent, methods.

The first method: The replacement of the water column by tho plexiglass rodintroduces a small positive reactivity which is proportional to the square of neutrondensity at the perturbed center according to the perturbation theory [5]. Theconstancy of the proportional coefficient is obtained for nil the investigatedpositions.

The second method: The perturbation may cause a strong change of the neutrondensity at the replaced position but no significant change at tho other places inthe core [4]. This.effect is also verified for tho considered positions.

The perturbation of the neutron field is checked for the replacement of thefuel element by tin? plexiglass rod using the second method.

From the experimental result it is concluded that the replacement of thomeasurement of relative distribution of neutron Held on the surface of fuel elementby that in the water column or the plexiglass rod la acceptable. This experimentalmethod will be used for the measurement of tho spatial distribution of the thermalneutrons presented in Section 6.

0. SPATIAL DISTRIBUTION OF THE THERMAL NEUTRON FIELDThe replacement of tho former TRIGA reactor core by tho new core with the

neutron trap at the center and the supplementary beryllium reflector layer at thoperiphery makes the active core to have a rim shape witli 4.3 cm inner radius,

226

5

w«

•

0.5

1 .0

0 . 5

1 . 0

0 . 5

-

.

t

*

f

r

to

1

¥

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*

4

• 1 ,5 #"

Position 7-3

PoattliHt V-v

•PualtitMi 1-4

4b

I

i

•

•Vi

•

4

-

*

-

20 20 JO 40 50 •(««)

- Anial d is l r ibu t lw ta o> tn» ll>«r««| n M t t u i t• t t»»« post tlo<»« 7 -3 , 7-» ami 1-4 twror- f 3X1• No I , « Mo 2 , U No 3 ,e Ho 4 , n Mo S, t No 4 ,1 Mo 7, D M o t , V Mo V, f Ho 10.

sa

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i , i .

1,4

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T

1

At tln> oont«r of tho ooru (II-0)

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HoJlnl dlolrlbutlon ol lltu Ihortm) iiaulifniain tho noli** ooro.

£ KxporiaentC&loulutlon tt tii« l-o-flroup noutrondLfCuolon theory, vxpionalun (1)

227

18.9 cm outer radius and GO cm height. So the thickness of the active core is L4.6cm that is equal to 4 + 5 times of the fuel element sizu. The active core issurrounded by the beryllium and graphite reflectors. The material above andunder the active core consists of aluminum and wnlnr. This narrow active corucontains 89 fuel elements, 7 control pxls and ,'J irradiation facilities. Thereforethe spatial distribution of the neutron field will be complicated.

The measurement of thermal neutron field distributions Is usually carried outat few watts power by activation of foils placed on the fuel element surface. Thisis difficult to apply for the Dalat reactor because of its high critical powers asmentioned in Section 4. As analysed in Section 5, the measurement may be carriedout by foil activation in a waU»r column or on the surface of a dummy elementmade of plexiglass instead of the fuel element. In the experiment the (Ju foils areirradiated at 2.5 kW in 15 minutos.

In Fig.7 are presented the axial distributions of the thermal neutrons at thepositions 7-3, 7-9 and 1-4 for some combinations of the control rods' depths. Kig.8shows that almost all the curves have nonsyinmetrical shape which can beexpressed analytically as follows [61:

<I»(z) = [A -I- C(z-30)] cosBz(z-30) (3)where A, C, Hz are constants, 0 = 0 for symmetrical curve and Bz is an axialbuckling.

Table 2. Neutron physics parameters of the Dalat reactor.

Parameteru

PowerFnol element, typoCritical coi-e loading

- without neutron trap- with noutron trap- operational core loading

Maximum thermal neutron llux, n/cin A*

Effective delayed neut.mu fraction, /io)f

/'oil"" '

lnhoinogeneous coefficients- radial- axial

Buckling, in'• radial- axial

Effective size, cm- radial- axial

Extrapolated distance), cm- radial- axial

lCxperiment.nl data

500UWVVR-M2

(59 fuel elements72 fuel elements89 fuel elements

1 *l2.1 IOM

(0 49±0.02)//(lf|-

I 77 iO 11uuiom

H'l.liiTi.n10.0-1:0.0

27()±l.O77.0±0.9

8.7±1.68.5±0.5

Calculated & TRIG Adata

TRICA: 250kWTRIG A: i;-7alln

77-80 [81, GO [10J77-80 181, 72 [10|

04 [41", 89 [10]2.28 10Kl f 101

1.8 l()i:i [8|0.81 \0'2 [HI

1 71 [H|, 1.74 [1()|1.3!) 18, 10]

84.0 [9]Hi.4 [91

26.2 [9]77.7 [9]

8.8 [9]8.8 [9]

a: CalcuUUion in Itef.[8J is carried out for the core Icxuled with 94 fuel elements.

The radial distribution of the thermal neutrons in the active core is shown inFig.8. It is obvious that the curve has maximum values nearby the neutron trap,

228

decreases rapidly within the first fuel element, after that decreases slowly andfinally raises nearby the beryllium reflector. This distribution can be expressedby the two-group neutron diffusion theory |7]:

<)>(r) - 14 JJa2(r-a3)J -I a4 I0|.H5(r-ii:l)] (4)where Jo and Io are the Besscl functions, aj -f- 115 are constants.

The constants in the expressions (3) and (4) are determined from experimentaldata by the last square fitting method. So the geometrical pa mine tors of the coresuch as innomogeneous coefficients of neutron field in axlat and radial directionskz and kr, axial and radial bucklings H2 and Br , effective height IIorr andeffective radius R ff, axial and radial extrapolated distances Xz and Av are obtained(Table 2). In Table 2 are also presented calculated values of the above - mentionedparameters [8, 9, ]0] which agree well with the experimental result.

7. DISTRIBUTION OF THE THERMAL NEUTRON AROUND THEGRAPHITE REFLECTOR

The active core of an equivalent radius of 18.9 cm is surrounded by the berylliumlayer of 2.9 cm average thickness and the graphite reflector of a rim shape with45.2 cm inner diameter and 34.(5 cm thickness. Inside the graphite block thereare three big holes coaxial with three horizontal beam tubes. A deep drain of 30cm depth and 7 cm width is made on the upper part of the graphite reflector forinstalling the rotary specimen rack which has 40 irradiation holes with diameterof 3 cm and depth of 20 cm in distance of 3f) cm from the active core center andoperates completely in water.

The defemination of the thermal neutron field distribution inside the graphitereflector is unable because of its solid structure. Therofore the measurement iscarried out by the irradiation of Cu foils at the inner and outer surfaces and onthe upper surface of the reflector. The axial distribution curves at the inner andouter surfaces have nonsymmetrical shapes. There is a difference betweon thesecurves but not significant. The radius distribution of thu thermal neutrons on theupper surface decreases with radius and has a girat value at the position of theirradiation water hole (Fig.9). The experimental dala may ln». described by thetwo-group diffusion neutron theory |7|. The calculation HIIOWH that the curve liasa maximum value at the position of about 22.5 cm in the boundary of tho berylliumlayer and the graphite reflector.

The distribution of the thermal neutrons around the rotary specimen rack isdetermined by the activation of Cu foils placed at 20 irradiation holes. Fig. 10shows the experimental result of <I>/<<1>>, where <<!>> is the average value of the20 measured data. From Fig. 10 it is seen that the;re is a nonsymmentry of thedistribution: the holes at the direction of thu 1 and 2 horizontal tubes have valuesgreater than the holes at the 3 and 4 horizontal tubes direction. Thenonsymmentrical value; is 1.27

8. CONCLUSION

The main result of the investigation of the neutron physics characteristics ofthe Dalat reactor is as follows:

229

0,3

* 0,2

) : IH - Vi Had in I d i s t r i b u t i o n of I lie I I I U I M U I i tcutlOMM onthe upper M u i l a c e of I liu ftiaphitu r i r f l c c l o i .

<• : Kxpur i m c n l u l 0 n \ u .: I 'uncl l«w« (4 ) .

\ 4),0910,03

/ ^ O . ^ t t O ,

l,0ttO,u>» " ° • ° " o,9t)!w,()'i / IlKan tubeDo am mUc V i .oa to .o*

Mo

J. The i« i« l n e u t r o n M u i c s around t h « ' 2 O I r r a d i a t i o nl iu les o f t l ie ro ta r j r a i i t c l n s n r a c k .

1.02 • - 0 . 0 3 : I r r a d i a t i o n h o l e w i t h Cu ( o i l• nd f e l a l i v c f l u » v a l u u a I I li« bo t o .

: I r r a d i a t i o n h o l e w i t h o u t any Cu f u l l .

230

- The effective fraction of delayed pholonoutrons induced by Be(y,n) reactionsis 0.49%/Jeff and the extraneous neutron source left after reactor shut down givesthe neutron flux of lO'VlO8 n/cm2/s.

- Tim lowest power levels of critical state.;; of Mm rend or are about f> W •:• (>()W due to the existence of the extraneous source, but power level must be chosengreater 500 W for neutron physics experiments In sustain tho stable critical'! tyof the reactor.

- Tho perturbation effect is obtained when a water column or a plexiglassrod is substituted for a fuel element. In consequence, the measurement of thorelative distribution of thermal neutrons in the reactor may bo carried out by theactivation of foils placed in a water column or on the surface of a solid plexiglassrod instead of on the surface of a fuel element.

- The relative axial and radial distributions of the thermal neutrons aremeasured and the geometrical parameters of the coro such as the i nhomogoncouscoefficients, the Ducklings, tho effective height and radius, the extrapolateddistances are obtained.

- The thermal neutron distributions are measured around the old graphitereflector.

9. ACKNOWLEDGEMENTS

The authors are indebted to Prof. Pham [)uy Hion, Drs. Vu Ilai Long, TrailHa Anli, Nguyen Tac Anh, Phain Ngoc Chuong and Engineers Ngo Phu Khang,Pham Van Lam for useful discussions. It is a pleasure to acknowledge tho effortsof the reactor staff, and the equipment assistance from the Nuclear PhysicsDepartment and Radioproteetion Department.

REFERENCES

.1. S.R. Koepin. Physics of Nuclear Kinetics.Addimm - Weslay Publishing Company, lu<:., I!H>5.

2. A.K. Krasin el. al.Proceedings of 2nd (ieimvn Confunince, PJ.,f>7 I, MJ.SH.

.1 1/ncn C. Sclmiid. Critical assemblies and I'endnr research.John Wiley & Sons, lnc.,1971.

4. A.D. (ialanin. Theory of heterogeneous reactor.Moscow, Atomizdat, 1971 (in Russian).

5. X.N. Feinberg, X.B. Sichov, V.B. Troiansky. Theory of nuclear reactors.Moscow, Atomizdat, 1978 (in Russian)

6. L.S. Tong and J. Weiman. Thermal Analysis of PWR. ANS 1970.7. R.I. Murray. Nuclear Reactor Physics.

Englewood Cliffs, N.J. Prentice-IIall, Inc.,H)r>7.8. Technical design for reconstruction and expansion of the Dalat research reactor.

Vol.3, 1979 (in Russian).9. Nguyen Pin we I>an. Calculation of the reactor physics parameters of the Dalat

nuclear research reactor. To be published.10. Nguyen Phuoc bin, I la Van Thong, Ngo Quang Huy, Do Quang Binh.

Calculation of fuel burn-up and fuel reloading for the Dalat nuclear researchreactor. To be published.

231