R. E. Ginna, Response to Requests for Additional ... · the ASME Code. However, WCAP-7769 assumed...

Maria KorsnickSite Vice President

Constellation EnergyGeneration Group

R.E. Ginna Nuclear Power Plant, LLC1503 Lake RoadOntario, NewYork 14519-9364

585.771.3494585.771.3943 Fax

[email protected]

December 9, 2005

U.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001

ATTENTION: Document Control Desk

SUBJECT: R.E. Ginna Nuclear Power PlantDocket No. 50-244

Response to Requests for Additional Information Regarding RevisedLoss-of-Coolant-Accident Analyses

This letter is in response to the October 28, 2005 "Request for Additional Information RegardingRevised Loss-of-Coolant-Accident Analyses, R. E. Ginna Nuclear Power Plant (TAC No.MC6860)". Ginna committed to provide responses within 45 days, but acknowledged in aDecember 18, 2005 public meeting on this subject that, due to the comprehensive level ofanalysis being performed, certain responses would require additional time. Attachment 1provides the majority of the responses to the October 28 request. The balance of theresponses, denoted as "Post-LOCA Long-Term Cooling" RAls #2, #3, and #5 will be submittedby January 16,2006.

These responses do not include any other regulatory commitments. If you have any questions,please contact George Wrobel at (585) 771-3535 or george.wrobel~constellation.com.

A-BD i

I ODO14'4

STATE OF NEW YORK :: TO WIT:

COUNTY OF WAYNE

I, Mary G. Korsnick, being duly sworn, state that I am Vice President - R.E. Ginna NuclearPower Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this responseon behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained inthis document are true and correct. To the extent that these statements are not based on mypersonal knowledge, they are based upon information provided by other Ginna LLC employeesand/or consultants. Such information has been reviewed in accordance with company practiceand I believe it to be reliable.

Subscribed and sworn before me, a Notary Public in and for the State of New York and Countyof ,i /w7 , - ,this 9 dayof U x 2005.

WITNESS my Hand and Notarial Seal: '/hO Af, 1 A gesNotary Public

SHARON L MLERNcoery Weli, StaW ot New YolkMy Commission Expires: Reafim No. 01M16017755

MO cgi^/i F O4 '1,2

Attachments

Cc: S. J. Collins, NRCP. D. Milano, NRCResident Inspector, NRC

Mr. Peter R. SmithNew York State Energy, Research, and Development Authority17 Columbia CircleAlbany, NY 12203-6399

Mr. Paul EddyNYS Department of Public Service3 Empire State Plaza, 10th FloorAlbany, NY 12223-1350

Attachment 1Response to October 28, 2005 Request for Additional Information

By letter dated April 29, 2005 (Agency wide Documents Access and Management System Accession No.ML051260239), R.E. Ginna Nuclear Power Plant, LLC (the licensee) submitted an application to amendthe technical specifications (TSs) for the R.E. Ginna Nuclear Power Plant. Specifically, the licenseeproposed changes that would reflect the revised analyses performed in support of the planned extendedpower uprate. To complete its review, by letter dated October 28, 2005, the NRC staff requested thefollowing information:

Effects Of Post-LOCA Analysis on Containment Sump pH

1. In order to complete its evaluation, the staff needs to review the general assumptions andcalculations used by the licensee to prove that the containment sump pH will be maintained above7 throughout the duration of the accident.

Describe the procedure utilized for calculating pH of the containment sump water during the 30day period after a loss-of-coolant accident (LOCA). If the calculations were performed manually,describe the methodology and provide sample calculations. If a computer code was used, providethe input to the code and the results calculated by it.

Response:

Plant design criteria require that the containment sump be maintained within a certain pH rangeduring post-accident operation. This ensures the long term availability of the Safety InjectionSystem (SIS), and prevents iodine from re-entering the containment atmosphere. The addition ofboron during Containment Spray and Safety Injection makes the initial sump pH in the acidicrange. To neutralize the sump pH, sodium hydroxide (NaOH) is added from the spray additivetank (SAT) via the Containment Spray System. Sump pH is a function of both boron and NaOHconcentrations.

In this evaluation, the pounds of boron and the pounds of NaOH in the sump is determined byconsidering both minimum and maximum delivered volumes and concentrations from borationsources and the SAT. The boration sources include the Refueling Water Storage Tank (RWST),Reactor Coolant System, and the SIS Accumulators. The pH is determined for each case usingboric acid/NaOH titration curves. Although a computer code is used to facilitate the calculation,the results can be easily verified with a hand calculation.

For example the computer code inputs with respect to the minimum calculated sump pH are asfollows:

Mode 1 RCS Mass (Ibm): 2.70165E5RCS boron concentration (ppm): 2069 (BOL Max.)RWST mass injected (Ibm): 2.47825E6 (Max.)RWST concentration (ppm): 3050 (Tech. Spec. Max.)Accumulators mass injected (lbm): 1.4218E5 (Max.)Accumulator concentration (ppm): 3050 (Tech. Spec. Max.)

1


SAT minimum injected volume (cu ft): 140.1

SAT NaOH concentration (wt percent): 30 (min.)

Based on the above inputs the code simply determines the average concentration of boron andNaOH in the sump post accident. Accordingly for the average boron concentration in the sump thecode adds the masses injected at each boron concentration and divides by the total sump mass.And for the average NaOH concentration in the sump the code divides the total pounds of NaOHinjected by the total sump mass. These results are used in conjunction with boric acid / NaOHtitration curves to determine sump pH.

For the mibpimum sump pH case the code analysis results are as follows:Sump borcn concentration (ppm): 2950Sump NaOH concentration (ppm): 1208SumppH: \ 7.8

Best-Estimate Large-Break LOCA (LBLOCA) Analysis

1. In order to show that the referenced, generically approved LOCA analysis methodologies applyspecifically to Ginna, provide a statement that confirms that Ginna LLC and its vendor haveongoing processes to assure that the ranges and values of the input parameters for the GinnaLOCA analysis conservatively bound the ranges and values of the as-operated plant parameters.

Response:

Both Ginna LLC and its analysis vendor (Westinghouse) have ongoing processes which ensurethat the values and ranges of the Best Estimate Large Break LOCA analysis inputs for peakcladding temperature and oxidation-sensitive parameters bound the values and ranges of the as-operated plant for those parameters.

The Ginna LLC process consists of documenting input parameters and rounding ranges along withthe bases for the values in a "DBCOR" document. The "DBCOR" becomes the new "AccidentAnalysis Assumptions" when EPU is implemented. Any modifications to Ginna are required tofollow the modification configuration control process. This process requires the effect of themodification on any Accident Analysis Assumptions to be addressed and resolved. Also, theInstrument Uncertainty program ensures the instrumentation can operate within the rangesassumed in the accident analysis (BELOCA).

2


2. If the plant-specific analyses are based on the model and/or analyses of any other plant, provide thejustification showing that the model or analyses applies to Ginna.

Response:

The Best Estimate LBLOCA analysis and associated model to support the Ginna EPU are bothGinna plant-specific.

Over-Pressure Protection - Safety Valve Canacity

1 . In its July 7 application with supporting documentation, descriptions of the provisions to addressover-pressure protection were included for Ginna when operating at the uprated power. The NRCstaff is reviewing continued sufficiency of the design margin of the safety valve capacity at theuprated power. The information provided in the application only addresses the change to thepressurizer safety valve lift setting and does not address the adequacy of the safety valve capacity.Although Table 5.2-1 in the Ginna Updated Safety Analysis Report does refer to the AmericanSociety of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section m,"Nuclear Vessels," 1965, the application does not provide details about analyses that were done atthe uprated power to demonstrate the adequate relief capacity and to show that sufficient designquantify margin remains.

Westinghouse Report WCAP-7769, Revision 1, "Topical Report Overpressure Protection forWestinghouse Pressurized Water Reactors," dated June 1972, does provide demonstration ofcompliance for Ginna with Article NM-7000, "Protection Against Overpressure," in Section III ofthe ASME Code. However, WCAP-7769 assumed that Ginna operating at 1518.5 megawattsthermal (MWt).

Provide the analysis results, determined using methods consistent with those in WCAP-7769(including credit for the second (or later) safety grade trip from the reactor protection system), toshow sufficiency margin in the design capacity for the Ginna pressurizer and steam line safetyvalves, with Ginna operating at the uprated power of 1775 MWt.

Response:

The Ginna EPU overpressure analyses are consistent with the requirements of SRP 5.2.2. SRP 5.2.2requires that the second safety grade reactor trip signal be credited for safety valve sizing calculations.This is consistent with the safety valve sizing procedure discussed in Section 2 of WCAP-7769.WCAP-7769 states, "For the sizing, main feedwater flow is maintained and no credit for reactor trip istaken." This analysis is typically performed prior to construction of the plant to provide a basis for thecapacity requirements for the safety valves and the requirement of SRP 5.2.2 provides a conservative basisfor the number and design of the valves.

However, WCAP-7769 goes on to say, "After determining the required safety valve relief capacities,

3


as described above, the loss of load transient is again analyzed for the case where main feedwater flow islost when steam flow to the turbine is lost. For this case, the bases for analysis are the same as describedabove except that credit is taken for Doppler feedback and appropriate reactor trip, other than direct reactortrip on turbine trip." This describes the analysis performed in Chapter 15 of the UFSAR which verifiesthat the overpressure limits are satisfied with the current/latest design.

The analyses performed in support of the Ginna EPU Program are not safety valve sizing calculations - nochanges are being made to the safety valves as a result of this uprating. The Loss of External ElectricalLoad / Turbine Trip analysis performed for the EPU Program, presented in Section 2.8.5.2.1, demonstratesthat the safety valves have adequate capacity to maintain peak primary pressure below 110% of designwhich satisfies the requirements of GDC-15. GDC-15 applies to "any condition of normal operation,including anticipated operational occurrences" which does not include a common mode failure of the firstsafety grale reactor trip signal.

The Loss Sf External Load / Turbine Trip RCS overpressure analysis is performed to demonstrate that, inthe event o4 a sudden loss of the secondary heat sink, the associated increase in reactor coolant systemtemperature does not result in overpressurization of the RCS system.

Small-Break LOCA (SBLOCA) Analysis

RAI #1: Provide the full set of transient parameters for the 1.5, 2, and 3-inch break sizes that includesthe following:

1.1 core power1.2 core inlet mass flowrate1.3 break mass flow rate1.4 break quality1.5 pressurizer pressure1.6 inner vessel or core two-phase level1.7 clad temperature at peak clad temperature (PCT) location1.8 steam temperature at hot spot1.9 heat transfer coefficient at hot spot1.10 injection mass flow rate vs time (pumped should be separate from accumulator injection)1.11 condensation~rate in cold legs1.12 void fractions in each core node versus time

Response:The following plots address SBLOCA RAI #1, requesting transient parameters for the 1.5-, 2.0-, and 3.0inchbreak sizes.

1.1 Figures 1.1-1 through 1.1-3 show the core powerforthe entire length of the transient for the 1.5-, 2.0-and 3.0-inch breaks.

4


Figure 1.1-1: 1.5-Inch Core Power (Btu/s)

RGE EPU. 10% SGTP. 25% AO, 1.5 inch Break, NO CS SIGNALCore Power (Btu/s)

710

610

co

0)o 510

0)10

4* 10

*1

. . . . . I . . . I I . . I . I I .. . . . . . - . . .

0 2000 4000

Time (s)8000 10000 12000

5


Figure 1.1-2: 2.0-Inch Core Power (Buds)

RGE EPU. 10% SGTP. 25% A0. 2-Inch Break. 650 SEC CS SIGNAL- Core Power (Btu/s)

7

10

0 I

CO

10

410 -I

Time (s)

6

Re e to October 28,2005 Request f Ad infO!

RGE EP1J lpo.'.., (tu/s;!re PO

10

7


1.2 Figures 1.2-1 through 1.2-3 show the core inlet mass flowrate for the 1.5-, 2.0- and 3.0-inch cases.

Figure 1.2-1: 1.5-Inch Core Inlet Mass Flowrate

RGE EPUP 10%WFL

SGTP, 25% A0O 1.5 inch Break. NO2 0 0 CORE INLET

CS SIGNAL

71K111 -

.150r -

E10000.

0

Ck,or

0U;: 5000

s c

~ I - - ._.0,, - .~1

..

-. I .I I .

0 2d0b 4OO 600Time (s)

8X0 1o0o0 12000

8



RGE EPUI 10% SGTP. 25% AO. 2-Inch Break, 650 SEC CS SIGNALWFL 2 0 0 CORE INLET

20000

15000-

C-'

E10000

.-4

0ir:

0

0

V)

O

I1

. I lIk..j. . L n ... ... I - . f- L

,, ,, .. ._ .! h 110IN _.I.1 -. -..-.-

I I I I I-MRJM -

2000 4000Time (s)

6000 8000

9



RGE EPU. 10% SGTP, 25% AO. 3-Inch Break- WFL 2 0 0 CORE INLET

20000

15000

CO,

E_10000-

I.-

0M 5000co

10


1.3 Figures 1.3-1 though 1.3-3 address the request for break mass flowrate for the 1.5-, 2.0- and 3.0-inchtransients.

Figure 1.3-1: 1.5-Inch Break Mass Flowrate

RGE EPU, 10%.- WFL

SGTP. 25% AO. 1.580 0

inch Break, NO CS SIGNAL0 BREAK TOTAL FLOW

400 -

300.

-PI

E(D

CO 200-

3:n

0

C',(j)

100 -

-I.-

K'I

I

0 2000 4000 6000 8000 10000 12 W0

Time (s)

11



RGE EPU, 10% SGTP, 25% AO. 2-Inch Break. 650 SEC CS SIGNALWFL 80 0 0 BREAK TOTAL FLOW

E 400

U 400

a:C,

12


Figure 13-3: 3.0-Inch Break Mass Flowrate

RGE EPU, 10% SGTP. 25% A0O 3-Inch BreakWFL 80 0 0 BREAK TOTAL FLOW

1600

1400

1200

C,)

E hooo

3C

4800

20

r, 600

0

0 1000 2000 30 4000Time (s)

13


1.4 Figures 1.4-1 through 1.4-3 show the break quality for the 1.5-, 2.0- and 3.0-inch break cases.

Figure 1.4-1: 1.5-Inch Break Quality

RGE EPU.

.8

.6

10% SGTP. 25% AO, 1.5 inch Break, NO CS SIGNALBreak Qual ity

0

(.'D

.4 -

.2 -A

0 , .. . , . . .

0 2000 4000 600o 8000 10000Time (s)

1200

14



RGE EPU. 10% SGTP. 25% AO, 2-Inch Break, 650 SEC CS SIGNALBreak Quality

.8

.6

.4-,

15



RGE EPU, 10% SGTP. 25% AO, 3-Inch Break- Break Quality

.6-

.4-

.2 -

0I0 1000 2000 3000 4000

Time (s)

16


Figures 1.5-1 though 1.5-3 show the pressurizer pressure transient for the 1.5-, 2.0- and 3.0-inch breaks.

Figure 1.5-1: 1.5-Inch Pressurizer Pressure .

RGE EPUI 10%PFN

SGTPR 25% AO. 1.59 0

inch Break, NO CS SIGNAL0 PRESSURIZER

2500

2000 1

.r5CL

4) 1500 -

co

1000-

am U I . . . . . . . . .t . . . 42000 4o00 600

Time (s)8600 10000 12000

17

Attachment 1Response to October 28, 2005 Request for Additional Infonnation

Figure 1.5-2: 2.0-Inch Pressurizer Pressure

RGE EPUP 10% SGTP. 25% AO. 2-Inch Break, 650 SEC CS SIGNALPFN 9 0 0 PRESSURIZER

2500 -

01_

,0

Time (s)

18



RGE EPUP 10% SGTP.PFN 9

2500

2000

1500

cL

CO,

500-

0O ' ' ' .

25%0

AO, 3-Inch Break0 PRESSURIZER

19


1.6 Figures 1.6-1 through 1.6-3 show the core mixture level for the 1.5-, 2.0- and 3.0-inch breaks.

Figure 1.6-1: 1.5-Inch Core Mixture Level

RGE EPU, 10% SGTP. 25% AO, 1.5 inch Break,EMIXSFN 7 0 0 CORE

- -- - TOP OF CORE = 20.3257 ft.

32 -

28 -

3-j

-6-' 24 -1

NO CS SIGNAL

20



RGE EPUi 10% SGTP. 25% AO. 2-Inch Break. 650 SEC CS SIGNALEMIXSFN 7 0 0 CORE

-- -- TOP OF CORE 20.3257 ft.

32.

30-

28

^26 -

24 -

~22

Time (s)

21



RGE EPU, 10% SGTP, 25% A0O 3-Inch BreakEMIXSFN 7 0 0 CORETOP OF CORE = 20.3257 ft.

32

ct 26

3 24

.-

x ,22

22


1.7 Figures 1.7-1 though 1.7-3 show the clad temperature at the peak clad temperature location for the entiretransient as well as the maximum local oxidation at the maximum local oxidation location for the 1.5-,2.0- and 3.0-inch break sizes.

Figure 1.7-1: 1.5-Inch Clad Temperature at PCT Elevation and Maximum Local Oxidation

RGE SBLOCA UPRATING 1.5-AB IN BRK NO CS SIGNALsblocta -23.0 Fl C2004/01/22 X2005/01/04

14:59:51.30 1021872120 bengalTemperature (F)

TCABY 31 0 0 ELEV 11.25Ox i de Thickness (7.)

-- ZIRCBY 31 0 0 ELEV 11.25

1100 .25E-01

I

1000

A2E-01

900 --.

.15E-01 co,

I.- I

800 - , , , , , , , , , , , , , , ,

E!a

a) I -

I .5E-02

500

400~ 0

23



RGE SBLOCA UPRATING ANALYSIS 2-AB IN BRK 650 SEC CS SIGNALsblocta 23.0 FP C2004/01/22 X2005/02/21

09 19:48.90 1966945675 siameseTemperature (F)

TCABY 32 0 0 ELEV 11.50Oxide Thickness (%)

ZIRCBY 31 0 0 ELEV 11.25

1\00. 01-O

-. 7-E-01

1000 - l E-01LZI900- .5E-01

800 -

4700 - -E-01

K 01

600 - .2E-01

400 I E01

24



RGE SBLOCA UPRATING ANALYSIS 3IN BRKsblocta 23.0 FP C2004/01/22 X2005/01/05

09:05:12.00 2062564891 siameseTemperature (F)

TCABY 30 0 0 ELEV 11.00Oxide Thickness (%)---- ZIRCBY 31 0 0 ELEV 11.25

1200 -. 25E-01

1100

.2E-011000

K IIj K o

900

0I .5E-0

C4

6-00 -

a ~.5E-02 .

600

500 0

25


1.8 Figures 1.8-1 though 1.8-3 show the steam temperature at the hot spot for the 1.5-, 2.0- and 3.0-inch break cases.

Figure 1.8-1: 1.5-Inch Steam Temperature at the Hot Spot

RGE SBLOCA UPRATING 1.5-AB IN BRK NO CS SIGNALsblocta 23.0 FP C2004/01/22 X2005/01/04

14:59:51.30 1021872120 bengal|.- TF 31 0 0 ELEV 11.25

P, 650I-

_.E

CL 600Ea,

I--

Time (s)

26



RGE SBLOCA UPRATING ANALYSIS 2-AB IN BRK 650 SEC CS SIGNALsblocta 23.0 . FP C2004/01/22 X2005/02/21

09:19:48.90 1966945675 siameseTF 32 0 0 ELEV 11.50

1000 -

900-

800 -

LU.

s 700-

Ea) --

27




09:05:12.00 2062564891 siameseTF 30 0 0 ELEV 11.00

700\

60 -CL.

E

28


1.9 Figures 1.9-1 though 1.9-3 show the heat transfer coefficient at the hot spot (PCT location) for the1.5-, 2.0- and 3.0-inch break cases.

Figure 1.9-1: 1.5-Inch Heat Transfer Coefficient

RGE SBLOCA UPRATING 1.5-AB IN BRK NO CS SIGNALsblocta 23.0 FP C2004/01/22 X2005/01/04

14:59:51.30 1021872120 bengalHB 31 0 0 ELEV 11.25

510

I10

C''

10

CL

co

10

I10 .

29



RGE SBLOCA UPRATING ANALYSIS 2-AB INsblocta 23.0 FP

09:19:48.90 1966945675 s- B 32 0 0

510 t

BRK 650 SEC CS SIGNALC2004/01/22 X2005/02/21

dameseELEV 1 1.50

CN j

L..

-.- 10e-

10

(0

I-)

(-10

D

W10

4000 5000Time (s)

30




09:05:12.00 2062564891 siameseHB 30 0 0 ELEV 11.00

510

104

L..10 -__'__'_'__'_'__'_'__'_,__'_'__'__'_,__'_'__'_'

C-4-

=3 3-4-10

10

10

Time (s)

31


1.10 Figures 1.10-1 through 1.10-3 addresstherequestforthesafetyinjection(SI)massflowrateversustimefor the 1.5-, 2.0- and 3.0-inch break transients. Figures 1.10-4 through 1.10-6 address the request foraccumulator injection mass flow rate versus time for the 1.5-, 2.0- and 3.0-inch break transients.

Figure 1.10-1: 1.5-Inch SI Mass Flow Rate

RGE EPU, 10% SGTPI 25% AO. 1.5 inch Break. NO CS SIGNAL(WFL 81 + WFL 82)

80 I

ED60-

Q)

0

40co,ci,0E

32



RGE EPU. 10% SGTP. 25% AO. 2-Inch Break. 650 SEC CS SIGNAL(WFL 81 + WFL 82)

100

80 :

E

60

40

Time (s)

33



RGE EPU, 10% SGTP. 25% AO. 3-Inch Break(WFL 81 + WFL 82)

100

E60

0E: 40

2000Time (s)

34


Figure 1.10-4: 1.5-Inch Accumulator Injection Mass Flow Rate

RGE EPU. 10%(WFL

SGTP. 25% AO.60 + WFL 61 )

1.5 inch Break. NO CS SIGNAL

120E

100 -

I-,

E--

0

0::

0cora

80 -

60-

40-

20 -

11

I 'i0 -I

I I. . I I I . I I I I I.

-LU -

6 2000 40D0 6000Time (s)

800 100o 12000

35



RGE EPU, 10% SGTP. 25% AO. 2-Inch Break, 650 SEC CS SIGNAL(WFL 60 + WFL 61)

50-

M'30

E

10

36



RGE EPU, 10% SGTP, 25% AO, 3-Inch Break(WFL 60 + WFL 61)

250

200

EE150

0

~100

2000Time (s)

37


1.11 Figures 1.11-1 though 1.11-3 show the condensation rate in the cold legs for the 1.5-, 2.0- and 3.0-inch break cases.

Figure 1.11-1: 1.5-Inch Condensation Rate in Cold Legs

RGE EPUP... ... ..

10% SGTP.WMIVFN 19WMIVFN 29

25% AO, 1.5 inch Break, NO CS SIGNAL

38



RGE EPU. 10% SGTP. 25% A0. 2-Inch Break. 650 SEC CS SIGNALWMIVFN 19

--- WMIVFN 29

5.

E-o

CD

ck,

C,,EX

Time (s)

39



RGE EPUS 10% SGTP, 25% A0. 3-Inch BreakWMIVFN 19

-_-- -WMIVFN 29

5.

0,

E

-o)

Time (s)

40


1.12 Figures 1.12-1 though 1.12-3 show the void fraction in each core node versus time for the 1.5-, 2.0-and 3.0-inch break transients.

Figure 1.12-1: 1.5-Inch Void Fraction in Core Nodes

RGE EPU,

.7 -

.6 -

C -

0

10-= >.4_ !*r 3 - 1.

10% SGTP3VFMFNVFMFNVFMFNVFMFN

25% A0O 1.53 04 05 06 0

inch Break, NO CS SIGNALO BOTTOM COREO LOWER-MID COREO UPPER-MID CORE0 TOP CORE

6000Time (s)

12000

41



RGE EPUP 10% SGTP.VFMFN

- VFMFN--VFMFN

-- VFMFN

.7-

.5 -

a

.) .4 -i

-0 .3 ,. jujl LII I

25% A0O3456

2-Inch Break. 650 SEC CS SIGNAL0 0 BOTTOM CORE0 0 LOWER-MID CORE0 0 UPPER-MID CORE0 0 TOP CORE

4000Time (s)

800

42

Attachment IResponse to October 28, 2005 Request for Additional Information


RGE EPU, 10% SGTPnVFMFN 3

---- VFMFN 4------- VFMFN 5--- VFMFN 6

_ - I

I.-I

::? .4 P i

25%0-000

AO. 3-Inch BreakO BOTTOM COREO LOWER-MID COREO UPPER-MID CORE0 TOP CORE

-Time (s)

43


RAI #2: What is the bottom elevation of the suction leg piping and the top elevation of the core? Also,provide the top elevation of the cold-leg discharge pipe.Response:

The bottom elevation of the suction leg piping is: 15.5336 ftThe top elevation of the core is: 20.3257 ftThe top elevation of the cold-leg discharge pipe is: 26.5108 ftNote: All elevations are with respect to the bottom of the reactor vessel.

RAI #3: Provide the limiting top peaked axial power shape used in the analysis.Response:

Figure 3- shows the hot rod limiting top peaked axial power shape used in the Reference 1analysis. e power shape takes into account 10% SGTP and 25% AO SBSH.

!

44


Figure 3-1: Hot Rod Limiting Power Shape

RGE Hot Rod Power Shape

16 8

14 -/

16

2-

4- _

0 , , , ,0_

Elevation (ft)

45


RAI #4:4.1 Provide the head flow curve for the pumped safety injection (SI) system for the severed emergency core

cooling injection line.4.2 Provide a set of plots for this break (see item 1).

Response:4.1 A severed emergency core cooling injection line is calculated as an 8.75-inch diameter break. Figure 4.1-

1 through 4.1-3 show the head flow curves for the pumped SI system for this break size.X Figure 4.1-1: RHR Flow Curve for 8.75-Inch Break

RHR Flow UPI (Analysis Input)

140 A --- _ _

1201 | - -- F 4- 4 + F

9 100

J 80

CM:

g h

- -. -_I=Ad4a

20

Av

0 20 40 60 100 120 140 16080

Pressure (psla)

46


Figure 4.1-2: Intact Loop HHSI Flow Curve for 8.75-Inch Break

Intact Loop HHSI Flows (Analysis Input)

45

40

35

30

25

W 20

IL 15

10

5

00 200 400 600 800 1000 1200 1400

Pressure (psia)

47

Attachment 1Response to October 28, 2005 Request for Additional InformationFigure 4.1-3: Broken Loop HHSI Flow Curve for 8.75-Inch Break

Broken Loop HHSI Flow - Spill to Containment (Analysis Input)

54.0

53.9 _ __ . __

53.7

i 53.6

* 53.5

3 53.4

53-53.3

L- -

531 __

0 200 400 600 800

Pressure (psla)1000 1200 1400

48

Attachment IResponse to October 28, 2005 Request for Additional MIformation

Response:4.2 Figure 4.2-1 through 4.2-12 show the set of plots requested under RAI #1 for the 8.75-inch break.

A 600 second SI interruption is assumed during switchover to recirculation for all SBLOCA casesanalyzed for R.E. Ginna. Note that this is a conservative assumption 'since it terminates SI flowscompletely during switchover. In reality, based on R.E. Ginna's EOP ES-1.3, the SI flows will not becompletely terminated during switchover. There will be flow from either the HHSI pumps or the LHSIpumps while the other pumps are being realigned should the RCS pressure allow. In particular, thisassumption would affect the larger break size cases, including the injection line break (8.75 inches),since the RCS pressure will be below the LHSI cut in pressure. As such, it would allow for continuousSI injection (either HHSI or LHSI) during the switchover period rather than the modeled 600 secondinterruption for the larger break sizes.

Note that in addition to the above assumption, for the injection line break case the NOTRUMP modelassumes the following:

1. Broken loop SI flow is spilled to the containment2. No flow is assumed from the broken loop accumulators. Therefore, Figure 4.2-9 shows the flow

from the intact loop accumulator only.3. No flow is assumed from the HHSI pumps following switchover to recirculation.

For the 8.75 inch break case minimal core uncovery is seen during this transient. Figure 4.2-12 showsthat the void fraction in the core reaches a value of 1.0 (indicating core dryout during that time) forvery short period of time during the initial blowdown period. This is short-lived in duration and has anegligible effect on PCT for this case.

Based on the above discussion, the rod heatup code was not run for the 8.75-inch break. Therefore, theplots requested in RAI #1 with respect to the peak clad temperature location and the hot spot are notincluded in this evaluation (Items 1.7 through 1.9).

49


Figure 4.2-1: 8.75-Inch Core Power (Btuls)

RGE EPU, 10% SGTP. 25% A0. 8.75-Inch BreakCore Power (Btu/s)

710

1,co~ 1

05C.) 10

410

Time (s)

50



RGE EPU. 10% SGTP,WFL 2

25%* 0

A0O0

8.75-Inch BreakCORE INLET

20000

15000

1-

E-Q

wv

0:0c

a:oa

10000

5000 -

0-

-5000 -

-10000 - I . I I . . . I I I i I i . I I 1 I I I I I I I I I-

I I I *

D 500 1000 1500Time (s)

2000 2500 3600

51



RGE EPU. 10% SGTP. 25% AO, 8.75-Inch BreakWFL 80 0 0 BREAK TOTAL FLOW

2000

E-o;

V

01000

Cx,

1500Time (s)

52



RGE EPU. 10% SGTP, 25% A0, 8.75-Inch BreakBreak Quality

I -

0.8 -

0.6

7F 7a

-_3

0.4 -

0.2

ao . . . ., . . . . I . . . . . . . . .

6 50 1600 1500 2000 2500Time (s)

A3000

53



RGE EPU. 10% SGTPR 25% AO, 8.75-Inch BreakPFN 9 0 0 PRESSURIZER

2500

2+0

.21500

a)

2- 1000

500 - -

0I

Time (s)

54



RGE EPUD 10% SGTP, 25% A0. 8.75-Inch BreakEMIXSFN 7 0 0 CORE

--- - TOP OF CORE - 20.3257 ft.

32

28-+

_ 26-

z_ 24 -

.x 22

ii

11"'1''. 1

I II11. -

III' II' I 1111 I I

-- - --- --- -- ---l- - --20 -

18 I

I I I I I I I I I I I , I .I I I I I . . . . . .

6 560 1000 1500Time (s)

2000 2500 3000

55


Figure 4.2-7: 8.75-Inch Broken Loop High Head SI Mass Flow Rate (Spill to Containment)

RGE EPU, 10%WFL

SGTP,81

25%0

A0O0

8.75-Inch BreakBL PUMPED SI

MIl -lvw

E

-aO

0

.20

10

. . . . .

la I I I I I I . . .. . . . . .

6 500 1000 1500Time (s)

2d00 l2500 3000

56


Figure 4.2-8: 8.75-Inch Intact Loop High Head SI Mass Flow Rate

RGE EPU. 10% SGTP. 25% AO, 8.75-Inch BreakWFL 82 0 0 IL PUMPED SI

V

40 -

U,

co

E

-0 0

20co

n10

L

I I I I U I I 1 1 I I I I I I U I a I I I _ I I I I I lU . ... .I.. .h. I

6 560 1000 1500Time (s)

2000 2500 3000

57


Figure 4.2-9: 8.75-Inch Intact Loop Low Head SI Mass Flow Rate

RGE EPU, 10% SGTP. 25% A0. 8.75-Inch BreakWFL 83 LHSI UPI Flow

140

E80--

060

CO,

An

58


Figure 4.2-10: 8.75-Inch Intact Loop Accumulator Injection Mass Flow Rate

RGE EPU. 10% SGTP. 25% A0. 8.75-Inch BreakWFL 61 0 0 IL ACCUMULATOR

1200,

1000-

c 800-E

0ck: 6003:

0 0o 400 -

200 -

0 I . I I , I I , I I I I I I I I I I I I I, I

'I.

6 560 Io 1500Time (s)

2000 2500 3600

59



RGE EPU, 10% SGTP, 25% A0. 8.75-Inch BreakWMIVFN 19

-- -- WMIVFN 29

10

\ 0

I

-0

0 _303:

co

as0Mn

I,'

I I I I 1 I 1 1 I I I l I I I I I I

-50-

-60 -

-70 - - -..

0 560 1000 150Time (s)

2000 2500 3000

60

Attachment

1



RGE EPUI 10% SGTPPVFMFN 3

- - - - VFMFN 4--- VFMFN 5- VFMFN 6

100 1

= 0.60

U-

-o>0.4

0.2

0500 1000

25%0.000

AO#0000

8.75 Inch BreakBOTTOM CORELOWER-MID COREUPPER-MID CORETOP CORE

I II I ,I I !W1500 2000 2500

Time (s)

L l 13000

---

61


RAI #5: Breaks larger than 3 inches were not provided. Provide information to demonstrate that breaks aslarge as 0.5 ft2 are not limiting and, as such, the worst-case break has been identified.

Response:

Break sizes larger than 3 inches in diameter were simulated using the NOTRULP computer code.These included cold leg breaks 4-, 6-, 8.75- and 9.75 inches in diameter. The 8.75 inch breakrepresents the severed injection line break and the 9.75 inch break represents the 0.5 ft2 break.

Figures 5-1 - 54 show the core mixture level plots for the above breaks. These plots show that there isno core ur~coveiy for the 4-and 6-inch breaks. Figure 5-3 and 54 also indicates that there is nosignificant sustained uncovery for the 8.75 and the 9.75-inch breaks. Therefore, for these cases it wasnot necessWay to perform a rod heat up calculation. Based on this these breaks are not limiting.Therefore itcan be concluded that breaks larger than 3 inches will not be limiting and the 2-inch breakcontinues to be the limiting small break case for R.E. Ginna.

62


Figure 5-1: 4.0-Inch Break Core Mixture Level

RGE EPUU 10% SGTP. 25% AO, 4-Inch BreakEMIXSFN 7 0 0 CORE

-- -- TOP OF CORE e 20.3257 ft.

32

30

28

2~26

M24

22

20 , . 300 .0 1000 2000 4000OW

63



RGE EPU, 10% SGTPR 25% A0O 6.0-Inch BreakEMIXSFN 7 0 0 CORETOP OF CORE = 20.3257 ft.

32

28

9 26

a)

64



RGE EPUS 10% SGTPR 25% A0O 8.75-Inch BreakEMIXSFN 7 O 0 CORE

--- - TOP OF CORE - 20.3257 ft.

32

30 - -

28

4--26

~22

lime (s)

65

Attachment IResponse to October 28, 2005 Request for Additional Information


RGE EPU. 10% SGTPR 25% AO. 9.75-Inch Break'EMIXSFN 7 0 0 CORE

--- - TOP OF CORE ' 20.3257 ft.

32 -

30\

28-

26

~24

66


Figure 5-5: 4.0-Inch Break RCS Pressure

RGE EPU, 10% SGTPR 25% A0. 4-Inch BreakPFN 9 0 0 PRESSURIZER

2500

2000

1500

C')P 1000-

67


Figure 5-6: 6-Inch Break RCS Pressure

RGE EPU, 10% SGTPR 25% A0O 6.0-Inch BreakPFN 9 0 0 PRESSURIZER

.0 1500

L..

CO

Ie 100D,(I.

Time (s)

68



RGE EPUP 10% SGTP, 25% AO. 8.75-Inch BreakPFN 9 0 0 PRESSURIZER

2500

2000

.rn0-2!

CO,P

ol

1500 4

1000-

I I I I I I I I I I

500 -

AnI I I I -L -L -L- I I I I lU - . I . . . I . . . I

0 510 10o0 150bTime (s)

200 25b0 3000

69

Attachment 1



RGE EPU, 10%PFN

SGTP.9

25%0

AO,0

9.75-Inch BreakPRESSUR I ZER

2500'

2000-

. 1500-

ram

C,

500

IU p I -. . . . I - . . - - . , . I . , , , I, . I.

- 560 Idoo 1T(sTime (s)

2000 2500 3000

70



RGE EPU.WFL

10% SGTP.80

25%0

AO. 4-Inch Break0 BREAK TOTAL FLOW

- 2000E

-o

a>0 1500

Q:

C,,Mo

0V 1000-

Time (s)

71



RGE EPU, 10% SGTP,WFL 80

7000 -

E _-o

4000

0

ca_

0 ,0

25%0

A0, 6.0-Inch Break0 BREAK TOTAL FLOW

72



RGE EPUP 10% SGTPoWFL 80

2000 -

1500- .

E

-0

(I)

c 1000-3:

at

25%0

AO, 8.75-Inch Break0 BREAK TOTAL FLOW

1500Time (s)

73



RGE EPU, 10% SGTP. 25% A0. 9.75-Inch BreakWFL 80 0 0 BREAK TOTAL FLOW

2000

1500

E

10SO - .

Dc) r

74


Figure 5-13: 4.0-Inch Break Quality

RGE EPU. 10% SGTP. 25% AO, 4-Inch BreakBreak Quality

so-~~ I Il..

Time (s)

75



RGE EPU. 10% SGTP, 25% A0O 6.0-Inch BreakBreak Qual ity

1;1l

76



RGE EPU. 10% SGTP. 25% A0O 8.75-Inch BreakBreak Qual i ty

.1 . ppIII*,*ip .1111111111 11.111.....1

0.8

0.6--

0.4-

0.2-

0 5 0 0 I I I I I I I I I0 500 1000

Ii

I30

* '1

I I I I I I I I I I I I I I I I

1500 2000 2500Time (s)

77



RGE EPU, 10% SGTP- 25,Break Quality

0.8

0.6 1 II

AO,

III III I

9.75-Inch Break

M111. Ii 11 I

V

~I '

I I I I II I I I I I I I I

0.4-

0.2 -

A

Ij

.. . . ..I * * * * * . * . * . . . .

u , .. . , .... , .... , , .... , ....

6 I~ io 1500Time (s)

2000 2500 3 l

78


Figure 5-17: 4.0-Inch Break Total SI Flow

RGE EPU, 10% SGTP. 25% A0, 4-Inch Break(WFL 81 + WFL 82)

100-

80

(I,E260

a)

40

20

0

Time (s)

79


Figure 5-18: 6.0-Inch Break Total SI Flow

RGE EPU, 10% SGTP. 25% AO, 6.0-Inch Break(WFL 81 + WFL 82)

100-

E-~60

al

20-

0

201000 2000 31d0 QTime (s)

80


Figure 5-19: 8.75-Inch Broken Loop High Head SI Mass Flow Rate (Spill to Containment)

RGE EPU, 10% SGTP. 25% A0O 8.75-Inch BreakWFL 81 0 0 BL PUMPED SI

Dlu_

50 -

40 -E

-o~

30301

0

c 20

10

I I I I . I I II, I IIII I I . J I .1 1. I . .U I~ .l

0 560 1000 1500Time (s)

2000 2500 3600

81


Figure 5-20: 8.75-Inch Intact Loop High Head SI Mass Flow Rate

RGE EPU. 10% SGTPR 25% AO, 8.75-Inch BreakWFL 82 0 0 IL PUMPED SI

50'

40

C,,

*o

10

0 * I I I I II I II I I I I I I I I f I I

Time (s)

82


Figure 5-21: 8.75-Inch Intact Loop Low Head SI Mass Flow Rate

RGE EPUI 10% SGTP. 25% AO, 8.75-Inch BreakWFL 83 LHSI UPI Flow

140

120 - -

100

E

80-

60

co0

0

83


Figure 5-22: 9.75-Inch Broken Loop High Head SI Mass Flow Rate (Spill to Containment)

RGE EPU. 10% SGTPR 25% A0O 9.75-Inch BreakWFL 81 0 0 BL PUMPED SI

w

50-

I-.. 40 -E

0

U)

101

(D

vK 30-

3-

ax

vc, 20M

10

, , ., | I . ,. . . I I | I I | { | 1I I | I lI I0 _ . . . . . . . . . . . . .I .

0 560 I60o I150Time (s)

200O 2500 3t00

84


Figure 5-23: 9.75-Inch Intact Loop High Head SI Mass Flow Rate

RGE EPU. 10% SGTP. 25% AO, 9.75-Inch BreakWFL 82 0 0 IL PUMPED SI

50M

40

E230

C: 20-

Cn

85


Figure 5-24: 9.75-Inch Intact Loop Low Head SI Mass Flow Rate

RGE EPU. 10% SGTP. 25% AO, 9.75-Inch BreakWFL 83 LHSI UPI Flow

140 -

120

100

80-

o60

(j)

An,

86


RAI #6:6.1 The 2-inch break shows an interruption in SI flow for about 500 seconds while the 3-inch break shows an

interruption of about 600 seconds.6.2 For the 3-inch break, the two-phase level in the upper plenum shows very little recession when the SI flow

has been terminated from 2700 to 3300 seconds. Explain and verify the insensitivity of the two-phase leveldue to the termination of SI flow. Provide an alternate means to verify core uncovery does not occur for allsmall breaks.

Response:6.1 The SI flow interruption for the 2- and 3-inch break cases were assumed at 600 seconds in the

NOTRUMP runs. Figures 6.1-1 and 6.1-2 present a close up view of the SI flows for the 2- and 3-inchbreak cases during the time of SI interruption. From these figures it can be seen that the 600 seconds SIflow interruption was correctly modeled for both the 2- and 3-inch breaks.

87


Figure 6.1-1: 2.0-Inch Break SI Mass Flow Rate

RGE EPU. 10% SGTP. 25% AO. 2-Inch Break. 650 SEC(WFL 81 + WFL 82)

CS SIGNAL

100

ci,

E

-0

a)

0

60 -

40 2

20 -

. I I I I I iS I I I I ~ I - II^, _

5000 5200 54b0 56oo 5800 6000Time (s)

88


Figure 6.1-2: 3.0-Inch Break SI Mass Flow Rate

RGE EPU. 10% SGTP, 25% A0. 3-Inch Break(WFL 81 + WFL 82)

100*

80

E260

Q:

40

20-

0

2400 2600 2800 300 3200 3400 3600Time (s),

89


Response:6.2 Figures 6.2-1 and 6.2-2 show the two-phase core mixture level for the 2- and 3-inch break cases.

These figures show that prior to the SI interruption, the two-phase mixture level is in the hot legregion. The mixture level begins to decrease when the SI flow interruption begins and continues todecrease to approximately the bottom of the hot legs when the switchover interruption ends after600 seconds. Figures 6.2-3 - 6.2-4a show the integrated flow between the hot legs and the upperplenum. From these figures it can be seen that there is backward flow from the hot legs into theupper plenum during the interruption period (Figures 6.2-3a and 6.2-4a). Figures 6.2-5 and 6.2-6show a comparison of the collapsed liquid levels in the active fuel and downcomer regions for the2- and 3-inch breaks. As can be seen, there is also considerable inventory available in thedowncomer region during the SI interruption period. These two factors would account for theminimal reduction in the core mixture level observed for the 3-inch break case.

90


Figure 6.2-1 2.0-Inch Break Mixture Level in the Core

RGE EPU, 10% SGTP, 25% AO. 2-Inch Break. 650 SEC CS SIGNALEMIXSFN 7 0 0 CORE

--- - TOP OF CORE = 20.3257 ft.… . BOTTOM OF HOT LEG = 24.1567 ft.

TOP OF HOT LEG = 26.5734 ft.

32 -

30 -

28-

26 -

22

91


Figure 6.2-2 3.0-Inch Break Mixture Level in the Core

RGE EPU, 10% SGTP. 25% AO, 3-Inch BreakEMIXSFN 7 0 0 CORE

- -- - TOP OF CORE = 20.3257 ft.…. BOTTOM OF HOT LEG = 24.1567 ft.-- TOP OF HOT LEG = 26.5734 ft.

32

6 26- -

9 24a)

.x 22L..

Time (s)

92


Figure 6.2-3 2.0-Inch Break Integrated Hot Leg Flow vs. SI Flow

RGE EPU, 10% SGTP. 25% AO. 2-Inch Break. 650 SEC CS SIGNALIntegrated Flow (Ibm)

Integrated Total Hot Leg FlowMoss Flow Rote (Ibm/s)

WFL 82 0 0 IL PUMPED SI

6000 - 50

4000-

_, ,_40

E I

-2000 * I~~ 0

93

Attachment 1Response to October 28, 2005 Request for Additional InformationFigure 6.2-3a 2.0-Inch Break Integrated Hot Leg Flow vs. SI Flow

RGE EPU. 10% SGTP. 25% AO. 2-InchIntegrated Flow (Ibm)

Integrated Total Hot LegMoss Flow Rate (Ibm/s)- - - -WFL 82 0

Break. 650 SEC CS SIGNAL

Flow

0 IL PUMPED SI

E-oBC 0

a)

Z-2000cm

a

- -E

ze

0-a:

co

C)

Time (s)

94


Figure 6.2-4 3.0-Inch Break Integrated Hot Leg Flow vs. SI Flow

RGE EPU, 10% SGTPR 25% A0OIntegrated Flow (Ibm)

Integrated Total Hot Leg FlowMass Flow Rate (Ibm/s)- - - -WFL 82 0 0 IL

3-Inch Break

PUMPED SI

EI--

0

L-o

-0

CP

50

40

0,

E30-

30_

20 ,en0)

10

0

95


Figure 6.2-4a 3.0-Inch Break Integrated Hot Leg Flow vs. SI Flow

RGE EPU, 10% SGTPR 25% AO,Integrated Flow (Ibm)

Integrated Total Hot Leg FlowMass Flow Rote (Ibm/s)_ - - - WFL 82 0 0 11

4000 -

2000

E-0

o0 IU-~

E -2000 lC '"

3-Inch Break

L PUMPED SI

E

a).4--

0

0CKi --

co

96


Figure 6.2-5 2.0-Inch Break Active Fuel RegionfDowncomer Collapsed Liquid Level

RGE EPU. 10% SGTP, 25% AO, 2-Inch Break. 650 SEC CS SIGNALActive Fuel Region Collapsed Liquid LevelDowncomer Mixture Level

- Downcomer Collapsed Level

.-

-

.5

-1

VCn

C-)

0 2000 4000 6000

Time (S)8000

uocbd - c=co on XZOW/11/07 by gagnc

97


Figure 6.2-6 3.0-Inch Break Active Fuel RegiorlDowncomer Collapsed Liquid Level

RGE EPU, 10% SGTP, 25% A0, 3-Inch BreakActive Fuel Region Collapsed Liquid Level

- - - -- Downcomer Mixture Level.--------- Downcomer Collapsed Level

35.

3-~

a 25-

-J

10-a 120 200D 300 400

Time (s)end on w X2005/11/17 by, g_

98


RAI #7: For larger breaks, the termination of SI flow is expected to have a more significant impact.Identify the impact of SI flow termination on the largest SBLOCA and the limiting LBLOCA.

Response:

The largest SBLOCA case simulated for R.E. Ginna is the 0.5 ft2 break. The results of the 0.5 ft2 breakare discussed in the response to question 5. In particular Figure 54 shows insignificant core uncoveryfor this case and hence no rod heat up calculation was necessary for this case. Therefore, it can beconcluded that the SI flow termination does not have a significant impact on the NOTRUMP results.

99


Post-LOCA Long-Tern Cooling

1. The long-term cooling analysis and boric acid precipitation analysis are based on a 1975Westinghouse letter that the NRC staff does not consider acceptable. Submit a new analysis thatcontains the following considerations for performing the long-term cooling analyses:

a. The mixing volume must be justified and the void fraction must be taken into accountwhen computing the boric acid concentration.

b. Since the mixing volume is a variable quantity that increases with time, the boric acidconcentration just prior to the switch to simultaneous injection should reflect the variablesize of the mixing region set by the pressure drop in the loop. The fluid static balancebetween the downcomer and inner vessel region (i.e. lower plenum, core, and upperplenum regions of the vessel) can then be performed taking into account the loop pressuredrop at a given steaming rate to compute the mixture volume in the core and eventuallythe upper plenum regions. The concentration in the resulting mixing volume just prior toexpansion into the upper plenum (which will cause a sudden decrease in concentration dueto the large area change in the upper plenum) must be shown to remain below theprecipitation limit.

c. The precipitation limit must be justified, especially if containment pressures greater than14.7 psia are assumed or additives are contained in the sump water.

d. The decay heat multiplier, as required by Appendix K to 10 CFR 50.46, must employ amultiplier of 1.2 for all times. 10CFR50.46(b)(5) states that "decay heat shall be removedfor an extended period of time required by the long lived radioactivity remaining in thecore." Appendix K, (I)(A) (4) entitled "Fission Product Decay" states that "the heatgeneration from radioactive decay of fission products shall be assumed to be equal to 1.2times the values for infinite operating time."

Response:

New boric acid analyses were performed that include the considerations listed above. Specificallytwo boric acid analyses were performed; large break LOCA and small break LOCA. A detaileddiscussion of results of these analyses will be presented in the response to RAI #2 and RAI #3.Below is a summary of the analyses relative to items "a" thru "d" considerations listed above.

For large hot leg breaks, cold leg injection will provide flushing flow during the injection mode.After re-alignment to sump recirculation cold leg injection will be terminated if system pressure isbelow UPI cut-in pressure. Once cold leg injection is terminated, boric acid will begin to build upin the core. The large break LOCA boric acid analysis determines the appropriate time forrestoring cold leg injection.

In regards to the "a" thru "d" considerations listed above, the following methodology was used for

100


the large break LOCA boric acid analysis.

a. Mixing volume / void fraction l core boiloff was extracted from WCOBRA/TRAC hot legbreak case.b. Mixing volume was varied with time as predicted by WCOBRAfTRAC. This mixingvolume reflects system effects such as loop resistance.c. The boric acid solubility limit was based on atmospheric conditions. Neither containmentoverpressure nor sump additives were credited.d. Appendix K decay heat was used in all calculations.

Small Break LOCA Boric Acid Analysis

For small 4cd leg breaks where the system pressure stays above UPI cut-in pressure, boric acidwill begin to build up in the core immediately after blowdown. Within 1 hour, operators willbegin to depressurize the reactor coolant system in accordance with Emergency Procedure ES-1.2,Post LOCA Cooldown and Depressurization. So long as the system can be depressurized prior toreaching the boric acid atmospheric solubility limit, there is no potential for boric acidprecipitation even with rapid inadvertent system depressurization using the pressurizerPORVs. The small break LOCA boric acid analysis determines how long the boric acidconcentration remnains under atmospheric solubility limit. This time is available for systemdepressurization.

In regards to the "a" thru "d" considerations listed above, the following methodology was used forthe small break LOCA boric acid analysis.

a. Mixing volume / void fraction / core boiloff was extracted from NOTRUMP 4 inch breakcase as this break requires no operator action to depressurize.b. Mixing volume was varied with time as predicted by NOTRUMP. This mixing volumereflects system effects such as loop resistance.c. The boric acid solubility limit was taken at atmospheric conditions. Neither containmentoverpressure nor sump additives were credited.d. Appendix K decay heat was used in all calculations.

2. Small breaks were not addressed. The boric acid concentration for the limiting SBLOCA needs tobe evaluated. Provide a summary of the results to show that the boric acid concentration is notsufficient to cause precipitation should the operators inadvertently depressurize the reactor coolantsystem (RCS) in a rapid manner.

Response:

As described in the cover letter, this request will be responded to by January 16, 2006.

101


3. Provide information to show that for the largest break that does not actuate upper plenum injection(UPI) (where a cooldown is required) that there is sufficient time to perform this function given anappropriate precipitation time based on consideration of the four items in item 1 above.

Response:


4. What is the temperature of the SI water entering the core at the time of SI re-initiation at 6 hours?

Response:

For large break and intermediate break LOCAs, cold leg SI would be terminated is at the end ofthe realignment to sump recirculation. Cold leg SI would be re-established after 6 hours, inaccordance with the Emergency Operating Procedures. At this point, the temperature of theinjected SI would be the temperature RHR heat exchanger outlet temperature. For small breakLOCAs, cold leg SI would not be terminated until the system is sufficient such that naturalcirculation will be established.

Regardless of the precise temperature of the SI injected into the cold legs, it is expected that thewater entering the core region via the lower plenum would be heated to near saturationtemperature. In addition to the stored energy in the metal structures in the downcomer (vesselwall, barrel, thermal shields or neutron pads), there are two other mechanisms to heat the injectedSI water as it travels to the core region. They are as follows;

a. Heat transfer from the core and upper plenum regions into the downcomer water through thecore-former structure or through the wall separating the upper downcomer from the upperplenum.

b. Steam flowing in the path from the upper plenum to the break will heat incoming water. Thiswould apply to water injected into either the hot legs or cold legs.

These mechanisms provide a means of heating incoming SI water making it highly unlikely thatwater in the lower plenum would be significantly below saturation temperature.

5. Once UPI initiates, at what time following a large break LOCA would the core steaming rate beinsufficient to entrain the hot-side injection?

Response:.


102


6. What are the guidelines for depressurizing the RCS below 140 psia? Describe the emergencyoperating procedure (EOP) requirements to accomplish this? Is there a time constraint forinitiating a cooldown? Does the cooldown consider a failure of the steam generator atmosphericdump valves (ADVs)?

Response:

Operators are trained and procedures are structured to rapidly depressurize the reactor coolantsystem during a LOCA. Emergency Procedure E-0, Reactor Trip or SI, will be implementedimmediately after the reactor trip. In this procedure operators verify operation of EngineeredSafety Features (ESF) equipment and diagnose the LOCA event. By experience in the simulator,after 10 mnutes operators will transition to Emergency Procedure E-1, Loss of Reactor orSecondary\ Coolant, where additional equipment is verified operational to enhance the cooldown(charging, instrument air, service water, etc.) and the need to cooldown and depressurize isidentified. Again, based on simulator experience, operators will transition to EmergencyProcedure ES-1.2, Post LOCA Cooldown and Depressurization, in approximately 30 minutes.While in ES-1.2 operators will commence the plant cooldown and depressurization. Experiencewith simulator training crews indicates that the RCS cooldown will be started within one hour ofthe break occurring. There is no time constraint for initiating a cooldown and one is not necessary.Operators are aware of the importance of depressurizing the RCS in order to stop the loss ofinventory and a time constraint for commencing a cooldown could be a distraction.

Emergency Operating Procedure ES-1.2 requires cooldown of the RCS at the maximum cooldownrate allowed by Technical Specifications. Operators are instructed to use both atmospheric dumpvalves (ADVs) to maximize the cooldown rate should the normal condenser dump valves not beavailable. Operators are also instructed to operate the ADVs locally if an active equipment failurecauses remote operation to fail.

7. If the RCS refills early during the cooldown for very small breaks and hot water is trapped in thepressurizer with a saturation temperature above the entry temperature to start RHR, how is thepressurizer eventually cooled to initiate residual heat removal (RHR)? Explain the method toreduce RCS pressure under these conditions. Can cooldown be accomplished before thecondensate storage tank supply is exhausted?

Response:

For very small breaks, the pressurizer is cooled using normal spray if reactor coolant pumps areoperating. If normal spray is not available, a PORV is opened to depressurize and cool thepressurizer. If no PORV is available, auxiliary spray would be used with charging pumpsavailable. Assuming none of the aforementioned methods are available, the Technical SupportCenter would be consulted and would likely recommend raising and lowering the pressurizer levelto mix the hot water in the pressurizer and the cooler water of the RCS. These actions are takenconcurrent with continuing the RCS cooldown by steaming from the steam generators.The capacity of one Condensate Storage Tank (CST) is adequate to maintain the RCS in hotstandby for two hours. Although two CSTs should be available, if only one tank is available it is

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possible that the water supply from that tank could be exhausted. In this case lake water is used asthe ultimate heat sink to supply makeup water using the standby auxiliary feedwater pump.

8. What precipitation limit is used for LBLOCAs and intermediate-break LOCAs? Explain whetherdebris in the sump water affects this limit and the time variation in boric acid concentration.

Response:

The boric acid precipitation limit used in both the large break LOCA and small break LOCA boricacid analyses is the experimentally determined solubility limit at the boiling point of asaturated boric acid and water solution at atmospheric conditions. The increased boric acidsolubility that would result from containment overpressure or the increased boric acid solubilitythat would result from sump additives was not credited.

While there are no known comprehensive industry studies of the effect of sump debris on boricacid precipitation characteristics, some relevant observations were made in the Fauske solubilitytests discussed in Fauske Report, FAII05-67, "Increase in Solubility Limit as a Result of SodiumHydroxide (NaOH) in the Containment Sump Water," dated June 2005. The Fauske test summaryreport indicated that powdered impurities introduced into a saturated boric acid water solution didnot cause boric acid to precipitate out of solution. Questions concerning this subject have beendiscussed with the NRC and the NRC has concurred that questions concerning this subject are notdirectly related to the uprating of the plant. The NRC has also indicated that an immediateresponse to these questions is not required for the safe operation of the plant. The low safetysignificance of questions related to boric acid precipitation is based on the following safetyassessment.

Safety Assessment of Generic Questions Regarding the Effect of Sump Debris on Post-LOCA BoricAcid Precipitation

During recent PWR Extended Power Uprating license amendment requests, discussions betweenthe NRC, licensees and safety analysis vendors have raised generic questions regarding the effectsof sump debris on the potential for boric acid precipitation after a LOCA. These questions do notconstitute a significant safety concern based on the following:

* There is low probability of a large break LOCA where conditions leading to significant boric acidaccumulation may be encountered. Small break LOCA scenarios are less likely to result in boricacid precipitation due to the higher boric acid solubility limit at higher pressures and the beneficialeffect of cooldown procedures on reducing core boiloff.

* Some of the transient behavior uncertainties that are treated conservatively, if fully understood andincorporated into boric acid precipitation analyses, would have a beneficial effect on the results.These include; liquid entrainment around the loop, boron carryover in the steam, containment

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overpressure, and mixing in regions outside the core, upper plenum and portions of the lowerplenum.

* If best estimate or realistic assumptions are used in the analyses, the predicted boric acidprecipitation times would be long after the typical EOP action times specified in plant's EOPs. Themost significant realistic assumptions are best estimate decay heat and nominal boron concentrationsfor containment sump boration and dilution sources.

* Observations from a test facility that represented a typical Westinghouse 3-loop PWR indicate thatboric acid precipitation would not occur for at least 24 hours after a large break LOCA (WCAP-16317-P, "Review and Evaluation of MI BACCHUS PWR Vessel Mixing Tests," November2004). These observations are indicative of the potential for boric acid precipitation in 2-Loop and4-Loop plant designs as well.

* The EOPs for\Westinghouse-designed PWRs provide the necessary steps and logic to implementactions that will mitigate the potential for boric acid precipitation in the core. EOP actions areparticularly relevant to small break LOCA scenarios where plant depressurization and cooldown is apriority. EOP actions would reduce or eliminate the potential for boric acid precipitation for manyof the scenarios considered in the boric acid analyses.

* In the event that boric acid precipitation should occur, it is unlikely that core cooling would becompromised. It is expected that the boric acid precipitate would tend to plate out on the colderstructures, accumulate in the lower plenum, or would collect at the top of liquid mixture level(Fauske Report FAI-05-13, "Solubility of Boric Acid in Water with TSP Added," 02-04-05 and CEReport LOCA-75-127, "Post LOCA Boric Acid Mixing Experiment," 10-06-75). It would takesignificant boric acid precipitate to totally block water from getting to the core.

9. For intermediate breaks that produce RCS pressures above the UPI shut-off head, the SI pumps aresecured if the need to switch to recirculation should occur. Explain the procedure for assuring RHRcan be actuated if the SGs have to be cooled down, especially with the loss of offsite power andfailure of one of the ADVs. What is the timing for cooldown of the SGs to assure RHR will beoperating when the switch to recirculation is made?

Response:

Operators will depressurize both steam generators using both ADVs. The single failure of anactive component is overcome by local manual control of the ADVs as necessary. A high-headrecirculation flow path is established upon entering the recirculation mode if the RCS pressure isabove the shutoff head of the RHR pumps. The high-head recirculation flow path involves liningup the discharge of the RHR pumps to the SI pump suction. This lineup produces adequateinjection pressure to assure continued injection flow at elevated RCS pressure.

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As previously stated, operators will maximize efforts to cool and depressurize the RCS. Givenhigh-head recirculation, it is not necessary to depressurize the RCS below the RHR pump shutoffat the time of recirculation in order to continue injection.

10. Following LBLOCAs, borated water is entrained in the steam exiting the core, which can enter theSG tubes. Since the secondary side is at high temperatures, the borated water can be boiled-offleaving behind the boric acid. What happens to the boric acid in the SGs? Can boric acid build-up sufficiently to increase the loop resistance and depress the two-phase level in the core?

Response:

There are no known industry studies of the potential for boric acid plating out on spacer grids, fuelalignment plate, or any structures in the upper plenum prior to the boric acid solubility limit beingreached. Once residual heat is removed from the non-heat-source structures (such as spacer gridsand fuel nozzles), there would be insufficient surface boiloff to create large amounts of depositionprior to hot leg switchover time. Concerning the possibility of plating out boric acid on the fuelrods, any plating on the fuel rods prior to fuel quench would likely be a boron compound otherthan orthoboric acid (discussed on page 220 of P. Cohen, P., 1969, Water Coolant Technology ofPower Reactors, Chapter 6, "Chemical Shim Control and pH Effect,") since the melting point oforthoboric acid is 3400F. After core quench, boron compounds that come out of solution might beexpected to return to solution quickly since locations below the mixture level would be exposed toa dilute core region solution. Boron compounds that return to solution and flow back into thevessel would be consistent with the boron acid assumptions used in the calculations (i.e. boric acidcontained in core boiloff remains in the core region).

Boric acid plate-out in the SG tubes could occur only during the period where there is significantliquid entrainment around the loop and only when the SGs act as a heat source. Under theseconditions, the rate at which boric acid accumulates in the core would be greatly reduced for tworeasons; the liquid entrainment passing through the SGs that is not vaporized would remove boricacid from the core region, and the boric acid plate-out on the inside of the tubes would beremoving boric acid from the core region. Once the SGs act as a heat sink, condensation on theinside of the tubes would return any boric acid to solution. Similarly, boric acid plate-out on otherstructures would return to solution when the residual heat in the structure is removed and theplated surfaces are exposed to a low-quality 2-phase mixture or liquid dilute solution. In responseto this RAI, a review of the Ginna large break boric acid analysis was made in order to estimate thevolume of boric acid that might be deposited in the SGs during the period when entrainmentaround the loop would be significant. Calculations have shown that 1 hr. 12 min. is the time afterwhich the core steaming rate is insufficient to support hot leg entrainment. At 1 hr. 12 min. a totalof about 1643 Ibm of boric acid would be left behind in the core as the result of boiloff. If it isassumed that 5% of that boric acid is entrained around the loops and left in the steam generators,the total boric acid per steam generator is 1643 * 0.05 / 2 = 41 Ibm. Assuming a density of boricacid of 50-100 Ibnuft3, the resulting volume per steam generator is less that 1 cubic foot. If the 1cubic foot of boric acid were to be deposited over a 10 ft length of steam generator tubing, thethickness would be 1 ft3 ((4765 tubes x 0.664 in I) / 12 in/ft x ii) * 10 ft) x 12 in/ft = 0.0014inches thick. This is not a sufficient deposition to cause a significant increase in loop resistance.

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11. Following a SBLOCA, the RCS can boil for an extended period of time. While the boric acid willremain in solution at the high temperatures, the sudden need to depressurize the RCS rapidly couldcause an inadvertent precipitation. Explain what guidelines or EOP directives are available to theoperators to assure this does not happen.

Response:

Analysis using the NOTRUMP computer code demonstrates that for smaller breaks the RCS willbe filled, iatural circulation will start and core boiling will cease before the boric acidconcentration exceeds the precipitation limit for atmospheric pressure. For larger breaks the RCSwill be de jressurized to the point where simultaneous injection will begin, also before the boricacid conceitration exceeds the precipitation limit for atmospheric pressure. Boron concentrationwill at no time exceed the saturation limit for a depressurized state and there is no concern thatrapid depressurization of the RCS could cause boron precipitation.

12. Explain how the EOPs guide the operators to assure them that they can refill the RCS for all smallbreaks and re-establish natural circulation to flush the boric acid from the vessel.

Response:

Operator training and procedures establish and maintain a high priority on depressurizing the RCS,returning normal level to the pressurizer and maintaining subcooling. If the break size is smallenough to support refilling the RCS, natural circulation will begin and core boiling and boronbuildup will cease. If the break size is too large to support refilling the RCS, the RCS will bedepressurized to the point where simultaneous injection will prevent boric acid buldup.

RAI#13: What is the size of the bottom mounted instrument tubes? Are failed instrument tubes in thebottom of the head part of the design basis? If so, was a failed tube analyzed at extended poweruprate conditions? Also, is operator action required to assure the core remains below 2200F withone tube assumed failed?

Response:

No plant specific analyses were performed for the R. E. Ginna Nuclear Power Plant (RGE) as part ofthe Extended Power Uprate (EPU) program with regards to failures of bottom mountedinstrumentation (BMI) tubes. However, in response to the Davis Besse and South Texas Unit 1events, a comprehensive Westinghouse Owners Group (WOG) program for both traditionalWestinghouse and Combustion Engineering System 80 reactor vessels was developed severalyears ago to assess the impacts of a postulated leak or failure of one or more BMJ nozzles. The WOGprogram included the following tasks.

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* Historical information review to determine the extent to which BMl breaks have been analyzed andto determine the effort required to address the potential consequence of a BMI failure.

* Small Break LOCA analyses to evaluate the potential effect of various failures of BMI tubes. Theseresults are then utilized to support a probabilistic risk assessment of BMI failures.

* A materials assessment which evaluates the potential for failure based on phenomenologicalconsiderations. This includes Failure Modes and Effects Analysis (FMEA)

* Evaluation of the effectiveness of the Emergency Response Guidelines (ERGs) in dealing with thispostulated scenarios and provisions for recommending modifications to the guidance.

During the execution of this program, various organizations discussed the benefits of providing acoordinated fleet-wide response to BMI related issues. As such, a joint effort between the WOG,B&W Owners Group (BWOG) and MRP was developed to provide this response. The effortculminated in the development of internal documentation which supports the various conclusionsreached in regards to these issues. A meeting to present the WOG and BWOG results to the NRC washeld on September 30 of this year. A summary of the observed LOCA response is provided below:

* Different plant groups demonstrate similar responses to the BMI small LOCA event. Evaluatedthermal hydraulic analysis cases representative of RGE show that a Bottom Mounted Nozzle (BMN)break of approximately 1.0 inch equivalent diameter can be withstood under timely operator action(45 minutes) to depressurize without core uncovery. (Note that the Ginna bottom-mountedinstrument tubes are 0.375 inches ID.)

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14. Explain the impact of the refilling of the loop seals (for breaks on the side of the cold leg) on themixing volume and boric acid concentration.

Response:

Loop seal refilling would temporarily increase the loop pressure drop and would depress themixture level in the core. Loop seal refilling would be significant to the calculations only if theloop seal closure was sustained. Neither LOCA ECCS Evaluation Models (EMs) nor observationsduring the ROSA tests (discussed in Letter from Westinghouse to NRC, NSD-NRC-97-5092,"Core Uncovery Due to Loop Seal Re-Plugging During Post-LOCA Recovery, March 1997)predict sustained loop seal closure, but instead predict cyclic loop seal refilling andclearing. Cyclic loop seal refilling/clearing would promote mixing in the vessel by forcing liquidfrom the core region to the lower plenum and downcomer. Effective mixing resulting from thistype of oscillatory behavior was observed in the modified VEERA facility tests (J. Tuunanen, etal., Experimental and analytical studies of boric acid concentrations in a VVER-440 reactor duringthe long-term cooling period of loss-of-coolant accidents, Nuclear Engineering and Design, Vol.148, 1994, pp. 217-231).

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