Crystal River Unit 3 - Response to Request for Additional ...

PDukeWEnergy®

Crystal River Nuclear PlantDocket No. 50-302Operating License No. DPR-72

September 6, 20123F0912-02

U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, DC 20555-0001

Subject: Crystal River Unit 3 - Response to Request for Additional Information to SupportNRC Reactor Systems Branch (SRXB) Technical Review of the CR-3 ExtendedPower Uprate LAR (TAC No. ME6527)

References: 1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - LicenseAmendment Request #309, Revision 0, Extended Power Uprate" (ADAMSAccession No. ML 112070659)

2. Email from S. Lingam (NRC) to D. Westcott (CR-3) dated July 23, 2012,"Crystal River EPU LAR - Draft RAIs from SRXB Associated with SpentFuel Storage (TAC No. ME6527)"

3. NRC to CR-3 letter dated August 3, 2012, "Crystal River Unit 3 NuclearGenerating Plant - Request For Additional Information For Extended PowerUprate License Amendment Request (TAC No. ME6527)" (ADAMSAccession No. ML 12213A303)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation (FPC) requested a license amendmentto increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts(MWt) to 3014 MWt (Reference 1). On July 23, 2012, via electronic mail, the NRC provided adraft request for additional information (RAI) related to spent fuel storage needed to support theSRXB technical review of the CR-3 Extended Power Uprate (EPU) License AmendmentRequest (LAR) (Reference 2). By teleconference on July 26, 2012, FPC discussed the draft RAIwith the NRC to confirm an understanding of the information being requested. On August 3,2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR(Reference 3).

The attachment, "Response to Request for Additional Information - Reactor Systems BranchTechnical Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI.

This correspondence contains no new regulatory commitments.

A-olCrystal River Nuclear Plant

15760 W. Powerline StreetCrystal River, FL 34428

U.S. Nuclear Regulatory Commission3F0912-02

of 3

If you have any questions regarding this submittal, please contact Mr. Dan Westcott,Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

/,Yon A. FrankeVice PresidentCrystal River Nuclear Plant

JAF/krw

Attachment: Response to Request for Additional Information - Reactor Systems BranchTechnical Review of the CR-3 EPU LAR

Enclosure CR-3 Spent Fuel Pool Boron Dilution Analysis

xc: NRR Project ManagerRegional Administrator, Region IISenior Resident InspectorState Contact

U.S. Nuclear Regulatory Commission of 33F0912-02

STATE OF FLORIDA

COUNTY OF CITRUS

Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida

Power Corporation; that he is authorized on the part of said company to sign and file with the

Nuclear Regulatory Commission the information attached hereto; and that all such statements

made and matters set forth therein are true and correct to the best of his knowledge, information,

and belief.

Jo A. Franke

ice PresidentCrystal River Nuclear Plant

The foregoing document was acknowledged before me this L04 day of

SJJ ,LA_, ,2012, by Jon A. Franke.

Signature of Notary PublicState of Florida

CARLYN E. PORTMANNCommission # DD 937553

. resMarch 1, 2014

(Print, type, or stamp CommissionedName of Notary Public)

Personally / ProducedKnown • -OR- Identification

FLORIDA POWER CORPORATION

CRYSTAL RIVER UNIT 3

DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72

ATTACHMENT

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION- REACTOR SYSTEMS BRANCH TECHNICAL REVIEW OF

THE CR-3 EPU LAR

U.S. Nuclear Regulatory Commission Attachment3F0912-02 Page 1 of 7

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -REACTOR SYSTEMS BRANCH TECHNICAL REVIEW OF THE CR-3

EPU LAR

By letter dated June 15, 2011, Florida Power Corporation (FPC) requested a license amendmentto increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts(MWt) to 3014 MWt (Reference 1). On July 23, 2012, via electronic mail, the NRC provided adraft request for additional information (RAI) related to spent fuel storage needed to support theReactor Systems Branch (SRXB) technical review of the CR-3 Extended Power Uprate (EPU)License Amendment Request (LAR). By teleconference on July 26, 2012, FPC discussed thedraft RAI with the NRC to confirm an understanding of the information being requested. Thefollowing provides the CR-3 formal response to the RAI needed to support the SRXB technicalreview of the CR-3 EPU LAR. For tracking purposes, each item related to this RAI is uniquelyidentified as SRXB X-Y, with X indicating the RAI set and Y indicating the sequential itemnumber.

1) (SRXB 1-1)

Section 2.8.6.2.2 of the technical report of the original EPU LAR dated June 15, 2011, states:

Studies performed in support of the evaluation discussed above demonstrate thatcontinued use of a uniform axial burnup profile remains conservative for EPUconditions.

For bumup credit applications, uniform axial profile becomes non-limiting after accumulatingsome amount of depletion. Explain what is meant by this statement and provide the "studies"used to demonstrate that the uniform profile remains conservative for EPU.

Response:

The statement regarding "Studies performed in support of the evaluation..." under thesubheading, "Results," in Section 2.8.6.2, "Spent Fuel Storage," of the CR-3 EPU TechnicalReport (TR) (Reference 1, Attachments 5 and 7), refers to previous studies performed for thecriticality analysis at pre-EPU conditions related to uniform (referred to herein as "flat") anddistributed axial fuel burnup profiles. These studies were confirmed for applicability to fueldepleted at EPU conditions. The studies, as previously documented in the CR-3 Spent Fuel PoolCriticality Analysis Report (Reference 3), determined that the flat axial fuel burnup profilebounds the distributed axial fuel burnup profile over the enrichment and burnup ranges as shownin Table 1, "CR-3 Pool A Axial Burnup Distribution Effect," of this attachment. These studieswere performed assuming fuel assemblies with axial blankets of 2.6 weight percent (wt%)enrichment of Uranium 235 (U-235) at the top and the bottom six inches of the active fuellength, which is conservative with respect to the current CR-3 fuel assembly design containing2.0 wt% U-235 enriched axial blankets. In addition, these studies were reviewed by the NRCstaff during the review of CR-3 LAR #292 that requested modification of the fuel storagepatterns in the CR-3 spent fuel pools. Based on this review, the NRC staff concluded that the flataxial fuel burnup profile is acceptable for evaluating fuel burnup greater than 30 gigawatt daysper metric ton unit (GWd!MTU) (Reference 2).

Additional calculations were performed to evaluate the impact of EPU operation on the axialburnup profile of a U-235 blanketed Mark-B-HTP fuel assembly. The fuel burnup results are


provided in Table 2, "Axial Burnup Profile for Blanketed Assemblies - EPU," of this attachmentand the comparison of the pre-EPU and EPU axial burnup profiles indicate minor deviations asshown in Figure 2.8.6.2-1, "Comparison of Pre-EPU and EPU Axial Burnup Distributions," ofSection 2.8.6.2 of the CR-3 EPU TR. A bounding distributed axial profile was calculated byselecting the minimum relative burnup of each node of the various assembly burnups from theplanned EPU core design. The results also indicate that the uniform axial profile continues tobound the distributed axial profile. The conclusion of these calculational studies is consistentwith the results and conclusion of the non-EPU spent fuel pool criticality analysis. As such, theflat (i.e., uniform) axial burnup profile continued to be used when evaluating the impact of EPUoperation on the CR-3 spent fuel pool criticality analysis.

Table 1: CR-3 Pool A Axial Burnup Distribution Effect

Enrichment 2.0 2.5 3.0 3.5 4.0 4.5 5.0(wt% U-235)Burnup (GWd/MTU) 6.41 13.26 19.56 25.47 31.94 37.46 42.73

Bumup Profile flat flat flat flat flat flat flat

k-calc 0.7801 0.7822 0.7832 0.7856 0.7810 0.7797 0.7805

stan dev 0.0006 0.0005 0.0005 0.0006 0.0006 0.0006 0.0006

Bumup Profile 40 40 40 40 40 40 40k-calc 0.7748 0.7770 0.7770 0.7771 0.7722 0.7700 0.7695stan dev 0.0006 0.0006 0.0005 0.0006 0.0005 0.0006 0.0005

Burnup Profile 50 50 50 50 50 50 50k-calc 0.7770 0.7789 0.7771 0.7761 0.7717 0.7710 0.7695stan dev 0.0005 0.0006 0.0006 0.0005 0.0005 0.0006 0.0006

Max 0.7801 0.7822 0.7832 0.7856 0.7810 0.7797 0.7805

Max Profile flat flat flat flat flat flat flat


AttachmentPage 3 of 7

Table 2: Axial Bumup Profile for Blanketed Assemblies - EPU

Axial Segment Relative Burnup(cm) Bounding Profile

0 to 15.2 0.2847.. .. 0 to .! ........L5 .2 .. ... .. ... 0 .84 ....... ........15.2 to i 25.06 0.7152

25.06 to 38.73 0.90992 5 0 _ ! t ... 8 7 . ............... 0 -9..9 ..............38.73 to 58.73 1.0309

.. . . . .. . . . . . . ..... . ... . . .. .. ........... . . ... ... .. .... ......... .. ... . ... ... ..... .. ..... .

58.73 to 78.73 1.081478.73 to 98.73 1.095498.73 to 118.73 1.0985118.73 to 138.73 1.099138.73 to 158.73 1.0992158.73 to 178.73 1.0997178.73 _ to_ 198.73 1.1008198.73 to 218.73 1.1023218.73 to 238.73 1.1039238.73 to 258.73 1.0998

258.73 to 278.73 1.0885278.73 to 298.73 1.0598298.73 to 318.73 0.9807 _

318.73 to 332.4 0.8141_ •3 ,3,,2 -4 ,,. i ..........4....2 ..................... 0.... 2_ _7.........

332.4 to 3422.6 0.6267342.26 to 357.46 0.2515

2) (SRXB 1-2)

Provide the spent fuel pool boron dilution analysisSpecification 4.3.1.

supporting the revised CR-3 Technical

Response:

Enclosure 1, "CR-3 Spent Fuel Pool Boron Dilution Analysis," to this attachment is provided toshow the time required to reach a spent fuel storage rack multiplication factor (ker) limit of 0.95in the CR-3 spent fuel pools during a boron dilution event. This analysis supports a CR-3 EPUlicensing basis change request to credit the use of soluble boron in the CR-3 spent fuel pools topreclude spent fuel pool criticality accidents as allowed by 10 CFR 50.68(b)(4). The analysisdetermined that the worst case credible spent fuel pool boron dilution event is the shear of a FireService System line and assumes a maximum fire water pump run-out flowrate of 3100 gpm.The analysis also assumes the initial spent fuel pool soluble boron concentration is at theminimum CR-3 Technical Specification limit of 1925 ppm boron and dilutes to a final solubleboron concentration of 571 ppm, which assures that the maximum kff is less than or equal to0.945 under accident conditions.

The analysis concludes that the time to reach a minimum allowable boron concentration of571 ppm is 77 minutes for Pool A, 60 minutes for Pool B, and 137 minutes when Pools A and Bare connected. Assuming a conservative time of 5 minutes for the spent fuel pool water level toreach the high level alarm, plant personnel have at least 55 minutes following receipt of the high


level alarm to terminate the event. Thus, it is concluded that, for credible dilution sources, thereis sufficient time for plant personnel to identify and terminate a boron dilution event prior toreaching the criticality limits required by 10 CFR 50.68(b)(4).

3) (SRXB 1-3)

Pool A racks contain Carborundum neutron absorber panels. Provide a detailed assessment ofthe current and future conditions of the Carborundum neutron absorber panel in terms of thefollowing:

a. uniform Boron- 10 loss from the absorber panels, and

b. local degradation such as gapping, cracking and/or scalloping.

Response:

Boron-10 Loss from Carborundum Absorber Panels

Since the installation of the Carborundum high density spent fuel racks in Pool A in early 1982,the spent fuel rack boron carbide (B4C) sample surveillances have been performed on sixoccasions. Various sample packets have been removed and B4C poison samples examined. Inaddition, during each refueling outage, the gamma sample holder is relocated, as directed by thereactor engineer, to the highest dose rate area of spent fuel Pool A to ensure the samplesexperience the highest dose to accelerate any degradation of the Carborundum samples.

Table 3, "CR-3 Spent Fuel Pool A Carborundum Sample Data," of this attachment provides theresults of the previous six B4C sample surveillances performed from 1984 to 2004. Thetable shows results for gamma exposed and water spent fuel rack samples. Each gamma andwater sample packet contains ten individual B4C samples positioned from the top (Sample 1) tobottom (Sample 10). Concentrating on the more limiting gamma exposed samples, nine of theten individual samples in the table show consistent results with Sample 2 being the outlier.Sample 2 has a vent hole, which causes damage to the individual sample directly adjacent to thehole. In 1998, Sample 2 in Packet 4 had the B4C and backing material missing, and in 2004Sample 2 in Packet 9 had a hole completely through the sample. It is unclear if some of thedamage occurred while the samples were in the gamma sample holder via flow induced erosion,or whether the damage was caused by the decontaminating process of spraying down the samplesupon removal from the sample holder. Regardless of the erosion mechanism, the damage hasbeen limited to the surface area directly adjacent to the sample packet vent holes, and has notbeen observed in the other samples above or below Sample 2. The actual rack vent holes arelocated approximately 8 inches above the active fuel height. As a result, Sample 2 is excludedfrom the total average weight loss column in Table 3 and a separate column is provided showingthe average weight loss associated with Sample 2.

Table 3 also shows a large percent loss over the first two years from 1982 to 1984, but greatlyreduced and consistent percent losses from 1984 to 2004 excluding Sample 2. This apparentlarge reduction in weight during this two year period was due to improperly establishing theinitial baseline density of the pre-exposed samples. During the first surveillance period, thevendor did not properly dry the samples to remove excess moisture prior to weighing the pre-exposed samples to establish an initial baseline density. Therefore, the actual pre-exposeddensities are unknown, and the measured weight loss over each subsequent surveillance period is


conservatively high compared to their pre-exposed weights. As a result, the more accuratemeasure of B 4C weight loss is shown from 1984 forward. Including the Sample 2 data and theconservative pre-exposed densities, the Carborundum neutron absorber panel sampling historyindicates approximately a 6% average B 4 C weight loss over the 22 year period.

Local Degradation of Carborundum Absorber Panels

During each high density spent fuel rack B4 C sample surveillance, each individual samplesurface is also examined for the appearance of texture, discoloration, cracking, scalloping,spalling (chipping), blisters, voids, and separation of the B4C granular surface from the fiberglassbacking. Excluding Sample 2, the other individual samples have shown consistent results withsome incidences of discoloration, but no major flaws.

As for swelling, the B4C sample surveillances do not include a thickness measurement; however,in reviewing the 2004 samples against the single CR-3 unexposed B4C sample, there is no visualindication of swelling through 2004. In addition, CR-3 has not experienced stuck assemblies orthe inability to insert assemblies in any cells of the high density spent fuel racks due to swellingas experienced at other facilities.

As for gapping, the Carborundum sheets used at CR-3 consist of one continuous piece versus theCarborundum panels that are stacked atop one another to form the full length panel. As such,individual panel shrinkage makes the panel type material more susceptible to gaps and not thesheet type material. For gapping to occur in the sheet type material requires cracking completelyseparating the individual sheet. To date, CR-3 has not experienced any cracking in the spent fuelrack sample.

In-situ Boron- 10 Areal Density Gauge for Evaluating Racks (BADGER) testing can reveal if anyof the Carborundum neutron absorber sheets have gaps or separations from cracking. Per theCR-3 license renewal application and associated RAI response letters, FPC has committed toBADGER testing as part of the enhancement of the Fuel Pool Rack Neutron AbsorberMonitoring Program, which will be implemented prior to CR-3 extended operation. Asdescribed in the FPC to NRC letter dated January 27, 2010, regarding an RAI associated with theCR-3 license renewal application (Reference 4), the BADGER surveillance test interval will beperformed every 10 years after the first BADGER test performance prior to CR-3 extendedoperation. The BADGER test will also be staggered with a 10 year interval for the B4C samplesurveillances, staggering them at 5 year intervals to ensure either B4C sample surveillances orBADGER testing is performed every 5 years.


AttachmentPage 6 of 7

Table 3: CR-3 Spent Fuel Pool A Carborundum Sample Data

Gamma SamplesSample Sample Sample Total Individual Sample Weight Losses Average Average Loss -

Removal Age Packet Gamma 1 2 3 4 5 6 7 8 9 10 Loss Sample 2Date (Years) Number Dose (Rad) (%) (%) (%) (%) (%) (%) (%) (%) (%) (%) (%) (%)

04/15/1982 0.002/17/1984 1.8 1 6.74E+09 2.670 2.870 2.520 2.260 2.290 2.350 2.250 2.230 2.290 2.370 2.410 2.35911/04/1985 3.6 10 l.OOE+10 2.201 2.669 2.367 2.109 2.214 2.118 2.220 2.419 2.195 2.098 2.261 2.21602/19/1988 5.8 2 1.43E+10 2.807 4.362 2.744 2.897 2.525 2.623 2.729 2.506 2.539 2.709 2.844 2.67505/14/1993 11.1 3 2.02E+10 4.040 10.250 3.530 6.230 3.720 3.590 3.650 3.640 3.600 3.740 4.599 3.97106/18/1998 16.2 4 3.32E+10 4.912 17.860 4.420 4.560 4.496 4.450 4.190 4.096 4.190 4.005 5.718 4.36905/20/2004 22.1 9 4.02E+10 5.537 21.010 5.066 4.263 4.259 4.329 4.365 4.331 4.740 4.491 6.239 4.598

Totals from 1988 to 1998 1.89E+10 2.105 13.498 1.676 1.663 1.971 1.827 1.461 1.590 1.651 1.296 2.874 1.693Totals Projected to 2036 (Extended Life) 1.01E+I 1 12.273 64.204 10.429 9.585 10.566 10.175 9.040 9.419 10.023 8.638 15.435 10.017

Water SamplesSample Sample Sample Total Individual Sample Weight Losses Average Average Loss -

Removal Age Packet Gamma 1 2 3 4 5 6 7 8 9 10 Loss Sample 2Date (Years) Number Dose (Rad) ( M) (%) M%) (%) (%) (%) (%) (%) (%) (%) (%) (%)

04/02/1982 0.002/17/1984 1.8 13 82.31 2.200 2.270 2.310 2.150 2.100 2.150 2.270 2.280 2.090 2.120 2.194 2.18611/04/1985 3.6 21 157.45 1.787 1.989 1.815 1.756 1.728 1.619 1.661 1.699 1.620 2.105 1.778 1.75402/19/1988 5.8 15 445.04 2.627 2.839 2.464 2.376 2.265 2.296 2.562 2.294 2.169 2.262 2.415 2.36805/14/1993 11.1 16 1590.63 3.850 3.940 3.510 3.150 2.710 2.930 3.290 2.860 2.680 3.010 3.193 3.11006/18/1998 16.2 18 1858.6 5.000 4.450 3.590 3.000 2.990 3.037 2.790 2.660 3.120 4.350 3.499 3.39305/20/2004 22.1 19 2040.8 5.414 4.970 2.708 2.221 2.235 2.394 2.637 2.908 3.168 4.578 3.323 3.140


4) (SRXB 1-4)

Provide a quantitative assessment of the impact of panel degradation mechanisms noted in RAI 3on the criticality analysis of record. Estimate the date on which the worst panel will exceed theassumed condition in the criticality analysis or record.

Response:

As shown in Table 3, "CR-3 Spent Fuel Pool A Carborundum Sample Data," of this attachment,B4 C sample surveillance periods from 1988 to 1998 represent the steepest rate of B4C weightloss over the 22 years. Table 3 shows the total accumulated dose along with the overallindividual and average B4 C weight loss for the ten year period. These totals exclude Sample 4from 1993, which, along with Sample 2, is an outlier compared to the other nine samples from1993 and the other Sample 4 surveillances throughout the 22 year period. Projecting this tenyear weight loss rate from 2004 to the end of the CR-3 period of extended operation, includingthe Sample 2 data and conservative pre-exposed densities, the approximate average weight lossthrough 2036 is expected to be approximately 15.5% average B4C sample weight loss with atotal accumulated gamma dose of approximately 1.OOE+l 1 rads. Based on previous vendortesting of this type of poison samples, a 20% weight loss of a B4C sample is equivalent to a 15%Boron-10 material degradation. Therefore the CR-3 spent fuel rack Carborundum absorberpanels are expected to remain within the 15% Boron-10 material degradation assumption of thespent fuel pool criticality analysis through the extended period of operation at the EPU powerlevel.

Based on the B4C individual sample surface examinations during the previous surveillances andover 22 years of exposure to CR-3 spent fuel pool conditions, the samples, with the exception ofSample 2, did not experience loss of B4C grains leaving appreciable voids, scalloping, orspalling. Also, the B4 C samples did not contain appreciable cracks, blisters, or separation of theB4C granular surface from the fiberglass backing. To date the Sample 2 damage has beenlimited to the surface area directly adjacent to the sample packet vent holes, and no suchdegradation has been observed in any of the other samples. The actual rack vent holes arelocated approximately 8 inches above the active fuel height, and if such deterioration wereoccurring in the racks, this Sample 2 degradation mechanism would have to be global and reachdown through the lower samples (3 through 10) to reach into the top of the active fuel region.Therefore, it is reasonable to conclude that these local degradation mechanisms will notadversely impact the capability of the Carborundum neutron absorber panels from performingtheir function and that the spent fuel pool criticality analysis will remain valid through theextended period of operation at the EPU power level.

In addition, the response to RAI B.2.37-2 associated with the CR-3 license renewal application,FPC to NRC letter dated January 27, 2010 (Reference 4), provides a description of how thematerial condition of the Carborundum neutron absorber panels will be monitored during theperiod of extended operation. To avoid duplication of NRC staff reviews, FPC proposes furtherquestions regarding the estimated schedule on when the degradation of the Carborundum neutronabsorber panels will exceed the assumed condition in the criticality analysis of record beincluded as part of the CR-3 license renewal application review.


References

1. FPC to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License AmendmentRequest #309, Revision 0, Extended Power Uprate." (ADAMS Accession No.MLI 12070659)

2. NRC to FPC letter dated October 25, 2007, "Crystal River Unit 3 - Issuance ofAmendment Regarding Fuel Storage Patterns in the Spent Fuel Pool (TAC No. MD3308)."(ADAMS Accession No. ML072910317)

3. Holtec Report HI-2063559, Revision 1, "Criticality Analysis of Additional Patterns forCrystal River 3 Pools A & B," dated September 19, 2006. (Proprietary)

4. FPC to NRC letter dated January 27, 2010, "Crystal River Unit 3 - Response to Requestfor Additional Information for the Review of the Crystal River Unit 3, Nuclear GeneratingPlant, License Renewal Application (TAC NO. ME0274) and Amendment #9." (ADAMSAccession No. ML100290366)

FLORIDA POWER CORPORATION

CRYSTAL RIVER UNIT 3

DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72

ENCLOSURE 1

CR-3 SPENT FUEL POOL BORON DILUTION ANALYSIS

SYSTEM#CALC. SUB-TYPEPRIORITY CODEQUALITY CLASS

SFN/A3S

NUCLEAR GENERATION GROUP

Fll-0001(Calculation #)

CR-3 SDent Fuel Pool Dilution Analysis(Title including structures, systems, components)

FII BNP UNIT

NCR3 [I] HNP FI]RNP I] NCP W-1ALL

APPROVAL M Electronically Approved

REV PREPARED BY REVIEWED BY SUPERVISORSignature Signature Signature

0 Name Name Name

Tyson Huntsman/ Lewis Wells Michael T. FloydRyan A. StephensDate Date Date

09/05/2012 09/05/2012 09/05/2012

(For Vendor Calculations)

Vendor Vendor Document No.

Owner's Review By Date

Calculation No. Fll-0001Page ii of ivRevision 0

TABLE OF CONTENTS

Table of Contents ii

Revision Summary iii

Document Indexing Table iv

Purpose 1

Assumptions 1

Design Inputs 1

Scenarios 4

Analysis 5

Results / Conclusions 6

References 9

Attachment 1 - Record of Lead Review 3 pages

Calculation No. F1 1-0001Page iii of ivRevision 0

Revision Summary

Rev. # Revision Summary (list ECs incorporated)

0 Initial issuance of this calculation in support of the Spent Fuel Pool CriticalityAnalysis for EPU conditions.

Calculation No. Fll-0001Page iv of ivRevision 0

Document Indexing Table

Document ID Number Function Relationship to Calc. ActionType

(e.g., Calc No., Dwg. (i.e. IN for design (e.g. design input, assumption basis, (specify if Doc. ServicesNo., Equip. Tag No., inputs or references; reference, document affected by(e.g. CALC, DWG, Procedure No., OUT for affected results) or Config. Mgt. to Add,

TAG, Sotaenm n ouet)Deleted or Retain) (e.g.,

PROCEDURE, Software name and documents) CM Add, DS Delete)SOFTWARE) version)

TAG FSP-1 IN DESIGN INPUT ADDTAG FSP-2A IN DESIGN INPUT ADDTAG FSP-2B IN DESIGN INPUT ADDPROCEDURE OP-406 IN REFERENCE ADDPROCEDURE OP-418 IN REFERENCE ADDPROCEDURE OP-880 IN REFERENCE ADDDWG 302-182 IN DESIGN INPUT ADDDWG 302-231 IN DESIGN INPUT ADDDWG 302-621 IN DESIGN INPUT ADDDWG 521-110 IN DESIGN INPUT ADDDWG 521-111 IN DESIGN INPUT ADDCALC 191-0006 IN DESIGN INPUT ADDCALC M98-0033 IN DESIGN INPUT ADDCALC M99-0008 IN DESIGN INPUT ADDCALC F97-0014 IN DESIGN INPUT ADDCALC HNP-M/MECH- IN DESIGN INPUT ADD

1099CALC M98-0055 IN DESIGN INPUT ADDOTHER ITS IN DESIGN INPUT ADDOTHER 00031-000 IN DESIGN INPUT ADD

I F

4 F F F

4 F I F

(For the purpose of creating cross references to documents in the Document ManagementSystem and equipment in the Equipment Data Bas

Calculation No. Fl1-0001Page 1 of 9Revision 0

1 Purpose

The purpose of this calculation is to identify and calculate the bounding dilution flow rates for theCrystal River 3 (CR3) Nuclear Plant spent fuel pools (SFPs), and perform a parametric study oftimes required to reach specified critical spent fuel pool boron concentrations. This analysissupports a request for a Technical Specification change which will credit boron concentration as amethod to maintain a sub-critical configuration in the A and B spent fuel pools as a result ofExtended Power Up-rate (EPU) at CR3. The analysis will be performed as directed by the KoppMemo (REF 7.4.4), and will consider all possible dilution initiating events (including operator error).It will also justify the surveillance interval for verifying the Technical Specification minimum poolboron concentration is maintained. The analysis will be performed such that it will apply to anyspent fuel pool configuration; 'A' and 'B' pools connected, or 'A' and 'B' pools separated by aphysical barrier.

2 Assumptions

2.1 SFP floor drains do not exist

To be conservative, it is assumed that none of the water from a pipe break is entering the floordrains and that the entire volume of water from the break is entering the SFP volume. Thisassumption will reduce the time required to reach the minimum critical boron concentration.

2.2 The boron concentration is homogenous throughout the spent fuel pool

To be conservative, it is assumed that all water mixes instantly, thereby neglecting additionaltime required for the pool volume to reach boron equilibrium.

2.3 The initial SFP level is at the low level alarm setpoint

To conservatively reduce the time necessary to reach minimum critical boron, it is assumedthat the spent fuel pool level is at the low level alarm setpoint. Plant procedures set aminimum spent fuel pool level 156' 6" ( REF 7.1.1) at which operator action is required.

3 Design Inputs

3.1 Nuclear Services Closed Cycle Cooling Water System (SW)

The Nuclear Services Closed Cycle Cooling Water System (SW) is an intermediate coolingsystem that removes heat from safety and non-safety related components during all plantoperation. The SW system is a closed loop system in order to help prevent direct leakage ofradiation from nuclear support systems in the plant to the environment. Though the SWsystem is not used directly to mitigate any accidents, it is used to provide cooling tocomponents in systems that are necessary to mitigate an accident.

Calculation No. Fll-0001Page 2 of 9Revision 0

Water from both the Spent Fuel Cooling System and the SW System flows through the SpentFuel Heat Exchangers (SFHE 1A & 1B). Table 3.1-1 shows the heat exchangercharacteristics for both the tube side (SF) and the shell side (SW) (REF 7.3.2, 7.3.3, and7.4.2).

Table 3.1-1SFHE 1A/1B Operating Characteristics

Tube Shell7.5 x 10A5 7.5 x 10A5

Mass Flow Rate Ibm/hr Ibm/hrPressure 77 Psig 189 PsigT in 127.2 Deg-F 95 Deg-FTOut 115.5 Deg-F 106.7 Deg-F

No. of Tubes 148

Tube OD 3/4inTube Thickness 20 BWG. Avg. *0.035 inTube Length 13.58 ft

*From Cameron Hydraulic Data

3.2 Fire Service System (FS)

The Fire Service System will be used as the bounding analysis for this calculation. Theanalysis will be based on operating the bounding 2000 GPM Fire Service pump at run-outconditions.

Pump run-out of the 2000 GPM FSP is 3100 GPM based on inspection of the pump curvelocated in REF 7.3.6. The pump discharge pressure of 220 ft at the flow rate of 3100 GPM isestimated to be equivalent to piping losses and elevation change for a pipe shear on the SFPbuilding floor, and is the maximum flow rate capability of the system.

3.3 Demineralized Water System (DW)

The DW system is used to provide demineralized water to the plant. It supplies demineralizedwater to the plant through the use of two identical 480V electric motor driven pumps. Each ofthese pumps is rated at 150 GPM and 300' total developed head (REF 7.3.7). For thepurpose of this analysis, the Demineralized Water System will be bounded by the 2000 GPMFire Service System pumps, and therefore will not be evaluated.

3.4 Spent Fuel Pool Design

The Fuel Handling Building consists of two spent fuel pools (Pools 'A' & 'B') that can be usedto store spent fuel. The two pools are connected by a canal that can also have a gateinserted to separate the two pools. Additionally, the B pool is connected to the Cask LoadingPit by another canal with a removable gate. These gates are used to allow draining of anindividual area of the pool without lowering the water level in the remainder of the system.Typical operation at Crystal River 3 has both gates removed so that the A and B pools and the


Cask Loading Pit are all connected. To be conservative for this calculation, analysis will bepreformed assuming that the cask pit gate is in place and it is isolated from the A and B pools.

The spent fuel pools and transfer canals are all built as a concrete structure with a stainlesssteel liner. The transfer canals are designed such that their bottom depth is above the top ofthe fuel racks. This is to prevent a loss of water in one section of the pool from uncovering thefuel in the other section of the pool if the removable gate is not in place. Table 3.4-1 wasgenerated by input data taken from CR3 calculation F97-0014, CR-3 Spent Fuel PoolTemperature Rise from Fuel in the Pool, and used to determine the total volumes of the spentfuel pool.

Table 3.4-1Total Raw Spent Fuel Pool Volumes

SSC L (N-S) Wume VolumeW) Det Voue Vlm

ft Ft ft Cu-ft gal

Pool'A' 23.917 32.125 38.166 29324.2 219360.4

Pool 'B' 23.917 32.500 38.166 29666.5 221921.0

Cask Pit 13.000 13.000 38.166 6450.1 48249.8CutoutPool 'B' 23216.5 173671.3

Total I I

Total A+B 52540.7 1 393031.7]

Table 3.4-1 gives the overall dimensions and volumes of the spent fuel pools (REF 7.3.4).The area included in Pool 'B' must have the area occupied by the 13' by 13' cask pitsubtracted to get the actual area of Pool 'B'. The Cask Loading Pit was not included in thetotal volume because it is assumed that the gate is in place, separating it from the rest of theSFP. Table 3.4-2 gives the partial Spent Fuel Pool volumes. These values are adjusted forspent fuel racks, spent fuel assemblies, and transfer tube platforms. For conservatism inTables 3.4-1and 3.4-2, it was assumed that the water level in the SFP is on the low levelalarm setpoint, resulting in reduced operator response times. Input data and detailedcalculations of the adjusted 'A' and 'B' spent fuel pool volumes can be found in calculationF97-0014, Revision 8. The adjusted volumes given in Table 3.4-2 will be used in the dilutionanalysis when determining boron dilution times.

Table 3.4-2Spent Fuel Pool VolumesTotal Adjusted

SSC Volume Volume

I Cu-ft gal

Pool 'A' 26602.0 198996.8Pool 'B' 20612.2 154190.0Total 47214.2 353186.7

Calculation No. Fl1-0001Page 4 of 9Revision 0

4 Scenarios

4.1 Break in the Spent Fuel Cooling System in heat exchangers

The Spent Fuel (SF) Cooling System consists of two independent and redundant coolingloops. The SF system interacts with the Service Water (SW) system through the Spent FuelHeat Exchangers (SFHEs) (REF 7.2.1). The heat exchangers have a tube and shell design inwhich a break in the tubes could result in leakage of unborated water into the spent fuel pools.The shell side has a design pressure of 200 psig. The tube side design pressure of 125 psig(REF 7.4.2). A break in a tube within the heat exchanger could result in an unborated waterleak into the spent fuel system due to the higher pressure on the shell side of the tubes. Thisscenario is bounded by the FSW line shear of 3100 GPM because the estimated max flow forthis scenario is 100 GPM.

4.2 Break in the Fire Service piping

The Fire Service System (FS), consists of three 2.5" standpipes and the associated hosereels (REF 7.2.3). One is located along the north wall, another along the east wall, and thelast along the southern most west wall. A break in one of these pipes is bounded by the 3100GPM Fire Service pump run-out analysis.

4.3 Break in Demineralized Water piping

Another possible source is the Demineralized Water System (DW). This system includes a 2"header running along the north side of the spent fuel pool. This header consists of one 1.5"valve and four 3/4" valves. Additionally, there are also another five individual 3/4" standpipes,each containing a single valve (REF 7.2.2). Four of these are along the east wall of the FHBand the last is along the southern most west wall. A break in one of these pipes is bounded bythe 3100 GPM Fire Service pump run-out analysis.

4.4 Misalignment of valve

There are a total of ten demineralized water valves that could be possible sources ofunborated water to the SFP if misaligned. The valve configuration of this system is asdescribed in scenario 3. There is one 1.5" valve that is a drain valve for the DW system. Thenine remaining 3/4" valves are all hose connection isolation valves. These hose connectionsare typically used to wash down the areas in the FHB. All ten of these DW valves arenormally administratively closed by REF 7.1.2. The Fire Service System consists of threeseparate 2.5" standpipes and their associated hose reels. Each of the three standpipes isreduced to an associated 1.5" hose reel isolation valve that is administratively sealed closedby REF 7.1.3. Misalignment of any of these valves is bounded by the 3100 GPM Fire Servicepump run-out analysis.

Calculation No. F1 1-0001Page 5 of 9Revision 0

4.5 Break in FHB roof storm drain

The FHB roof storm drains are an additional possible source of dilution. Because these stormdrains are only gravity fed, a break in these lines in considered highly unlikely. As a result, abreak in a storm drain will not be considered as a credible source of SFP boron dilution.

4.6 Tank ruptures in the vicinity of the pool

All large capacity tanks at the Crystal River 3 Nuclear Plant are located outside of the areathat could affect the SFP. Therefore, a tank rupture is not considered to be a credible dilutionsource to the SFP.

4.7 Dilution events initiated in the Reactor Coolant System

There are two (2) transfer tubes that separate the Reactor Building (RB) Fuel Transfer Canal(FTC) from the spent fuel pool. Each of these tubes is designed with a gate valve and a blindflange to separate the FTC from the SFP when not in refueling operations. During the courseof a refueling outage these valves may be opened for 3-4 weeks to permit fuel transferbetween the spent fuel pool and the reactor vessel. During this time, there is the possibility ofa dilution of the Reactor Coolant System (RCS) that would directly affect the SFP. TechnicalSpecification 3.9.1 requires RCS boron concentration to be maintained above the CoreOperating Limits Report (COLR) limit during Mode 6. Procedure FP-203, Offloading andRefueling Operations, administratively requires caution tags to be issued for applicable valvesthat may cause a dilution of the RCS during Mode 6. The possibility of dilution of the SFP dueto a dilution of the RCS is considered to be minimal and is therefore not analyzed.

5 Analysis

All scenarios described in Section 4 are either bounded by the pump run-out analysis for theFire Service System given in Section 3.2, or they are not considered to be credible dilutionsources. The bounding analysis is estimated to be conservative by approximately a factor of10 to the estimated worst case realistic SFP dilution flow rate of approximately 300 GPM.

5.1 Pump Run-out

Bounding analysis will be performed by determining the flow rate of water from a single 2000GPM FS pump at run-out conditions introduced into the Spent Fuel Pool. Run-out conditionsare determined using the pump curve for the FS pumps. Fire Service pump curves are foundin Attachment A of calculation M98-0055, Fire Service Design Pressure and Temperature, andshow a flow rate of 3100 GPM.

Calculation No. F1 1-0001Page 6 of 9Revision 0

5.2 Boron Dilution

The boron dilution rate for a constant volume is based on the following differential equation(REF 7.3.5):

Equation 5.1

Vdc- = -QC + QCl,dt

Where:C= boron concentration of mixing volume (ppm)Q= flowrate of water into mixing volume (GPM)V= finite mixing volume (gallons)Cin= boron concentration of water entering mixing volume (ppm)

This equation assumes the boron concentration remains homogenous throughout the volume.It can also conservatively be assumed that the boron concentration of the water entering themixing volume is equal to zero, thereby removing the QCin term from the equation. SolvingEquation 5.1 above for time (t) yields Equation 5.2 below.

Equation 5.2

Where C, is the initial boron concentration of the mixing volume at t=0. The final boronconcentration, Cf, is the final concentration desired in the mixed volume. When determiningtime for operator response, ratios of the final to initial boron concentration can be used withinthe formula. This will allow the calculation results to be used for a variety of initial and finalboron concentrations.

6 ResultslConclusions

6.1 Pump Run-out

Pump run-out for the Fire Service pump was determined from calculation M98-0055, FireService Design Pressure and Temperature. Attachment A of M98-0055 contains pumpperformance curves for the three 2000 GPM pumps. Run-out conditions correspond to thelast point on the pump curves of 3100 GPM.

6.2 Boron Dilution

Boron dilution analysis is completed using Equation 5.2 and the known spent fuel poolvolumes and flow rate for the FS pump run-out. Table 6.2-1 was generated based on ratios ofinitial (C, ) and final (Cf) boron concentrations.

Calculation No. F1l1-0001Page 7 of 9Revision 0

Table 6.2-1Boron Dilution Analysis

Pool A Pool B Pool A &B

Cf/Co connected

min min min

0.1 148 115 262

0.2 103 80 1830.3 77 60 1370.4 59 46 1040.5 44 34 790.6 33 25 580.7 23 18 410.8 14 11 250.9 7 5 121.0 0 0 0

The Technical Specification boron concentration limit is 1925 ppmB. The minimum allowableboron concentration to maintain sub-criticality in the SFP is identified as 571 ppmB by REF7.4.3. These boron concentrations yield a limiting Cf/Co ratio of 571/1925 or 0.296.

6.3 Spent Fuel Pool Fill Times

The time required to reach the high level alarm setpoint of the SFP can be determined basedon the flow rates of water into the SFP. Per REF 7.3.1, the Hi-Level Alarm setpoint is 159' 0".The required volume change in the pool to reach the Hi-Level Alarm point can be determinedby multiplying the difference in elevations, 2.5 feet, by the area of the pool in the givenconfiguration. Given this required volume change, necessary operator response time due toactuation of a Hi-Level Alarm can be determined for given flow rates into the SFP. All times toreach the Hi-Level Alarm have been rounded up to the nearest minute to be conservative.

The area of Pool 'A' is Length (23.92 ft) x Width (32.125 ft), or 768.3 Sq-ft. This results in avolume addition of 1920.8 Cu-ft or 14,369 gallons in order to reach the Hi-Level Alarm. Forthis volume and a flow rate of 3100 GPM, it would take 5 minutes before the alarm is receivedby the operators.

To create a limiting scenario for the area of Pool 'B', the required fill volume is determined byincluding the area of the cask pit and cask pit gate as well. The area of Pool 'B' is Length(23.92 ft) x Width (32.5 ft) - Cask Pit cutout (13' x 13') + Cask Pit (10' x 10') + Cask Pit Gate(3' x 3'), or 717.2 Sq-ft (REF 7.2.4 and 7.2.5). This results in a volume addition of 1793.2 Cu-ft or 13,415 gallons in order to reach the Hi-Level Alarm. For this volume and a flow rate of3100 GPM, it would take 5 minutes before the alarm is received by the operators.

The area with Pool 'A' & 'B' connected will be made conservative by also adding the area ofthe transfer canal between the two pools as well as the area of the cask loading pit and it'stransfer canal. This will result in the maximum time to reach the Hi-Level Alarm. The 'A' to 'B'canal is 4' by 3', the cask pit to 'B' canal is 3' by 3', and the cask loading pit is 10' by 10' (REF


7.2.4 and 7.2.5). Adding these areas to the already determined areas of the 'A' and 'B' poolsyields a total area of 1,497.6 Sq-ft. This results in a required volume change of 3,744 Cu-ft, or28,007 gallons. For this volume and a flow rate of 3100 GPM, it would take 10 minutes beforethe alarm is received by the operators.

The bounding scenario to reach the Hi-Level alarm is 10 minutes when Pools 'A' and 'B' areconnected and 5 minutes when they are isolated. This is the maximum time it would take foroperators to realize they have a leakage source into the SFP. At this time, operators wouldtake action to determine the source of in-leakage and prevent further dilution of the SFP. Theminimum boron concentration required to maintain sub-criticality is identified in REF 7.4.3 as571 ppmB. Using this limiting boron concentration, and a starting boron concentration at theTechnical Specification limit of 1925 ppmB (REF 7.4.1) yields Table 6.3-1 below.

Table 6.3-1Comparison of Alarm Time to Time to Reach Minimum Boron

Time to ReachTime to Minimum

Pool Alarm AllowableConfiguration Boron

min min

Pool 'A' 5 77

Pool'B' 5 60

Pool 'A + B' 10 137

Table 6.3-1 shows that for the worst case, plant personnel will have 55 minutes followingnotification of a dilution event to mitigate the effects of that event.

As a result of the bounding analysis performed in this calculation, it can be concluded that forall possible sources of spent fuel pool boron dilution, there will be sufficient warning to plantpersonnel of a dilution event before the minimum allowable boron concentration of 571 ppmB(REF 7.4.3) is reached. The Technical Specification surveillance interval of seven days forverifying the minimum SFP boron of 1925 ppmB is maintained is sufficient because a dilutionevent cannot occur without generating a Hi-Level Alarm and notifying operators.


7 References

7.1 Procedures

7.1.1 OP-406 Rev. 89, Spent Fuel Cooling System

7.1.2 OP-418 Rev. 48, Demineralized Water System

7.1.3 OP-880 Rev. 66, Fire Service System

7.2 Drawings

7.2.1 302-621 Rev. 54, Spent Fuel Cooling

7.2.2 302-182 Rev. 64, Condensate and Demineralized Water Supply for Nuclear Plant

7.2.3 302-231 Rev. 91, Fire Service Water

7.2.4 521-110 Rev. 10, Spent Fuel Pit - Floor Details

7.2.5 521-111 Rev. 05, Spent Fuel Pit - Liner Details

7.3 Calculations

7.3.1 191-0006 Rev. 1, Spent Fuel Storage Pool A&B Level Loop Accuracy

7.3.2 M98-0033 Rev. 2, Spent Fuel Cooling (SF) Design Pressure and Temperature

7.3.3 M99-0008 Rev. 0, Nuclear Services Closed Cycle System (SW) Design Pressure andTemperature

7.3.4 F97-0014 Rev. 8, CR-3 Spent Fuel Pool Temperature Rise from Fuel in the Pool

7.3.5 HNP-M/MECH-1099 Rev. 0, Spent Fuel Pool Boron Dilution Analysis

7.3.6 M98-0055 Rev. 0, Fire Service Design Pressure and Temperature

7.3.7 M93-0021 Rev. 2, Auxiliary Building Demineralized Water Storage Tank ( DWT-1)Volume

7.4 Other Documents

7.4.1 ITS, Improved Technical Specifications

7.4.2 00031-000 Rev. 2, Installation, Operation & Maintenance of Shell and Tube HeatExchangers

7.4.3 Holtec Report, HI-2063559 Rev. 2, Criticality Analysis of Additional Patterns for CrystalRiver 3 Pools A & B.

7.4.4 NRC Memo, Guidance on the Regulatory Requirements for Criticality Analysis of FuelStorage at Light-Water Reactor Power Plants, Acc. # ML072710248.

Calculation No. F1l1-0001Attachment 1

Page 1 of 3Revision 0

Record of Lead Review

Document F11-0001 Revision 000

The signature below of the Lead Reviewer records that:- the review indicated below has been performed by the Lead Reviewer;- appropriate reviews were performed and errors/deficiencies (for all reviews performed)

have been resolved and these records are included in the design package;- the review was performed in accordance with EGR-NGGC-0003.

- Design Verification ReviewL1 Design ReviewF-1 Alternate CalculationMi Qualification Testing

E-- Engineering Review El Owner's Review

Ej Special Engineering Review_______________________L-I YES L-I N/A Other Records are attached.

I ,Anwic WaIlkII Safety Analysis 09/04/2012

Lead Reviewer (print/sign) Discipline Date

ItemNo.

Deficiency Resolution

-q1 Need to reference the Kopp memo

(probably under "Purpose" and addressevery aspect of what is says should beconsidered:

"The analysis should consider all possibledilution initiating events (including operatorerror), dilution sources, dilution flow rates,boration sources, instrumentation,administrative procedures, and piping. Thisanalysis should justify the surveillanceinterval for verifying the technicalspecification minimum pool boronconcentration."

How is the surveillance interval addressedand justified?

Reference was added to the purpose.

Surveillance interval was justified because adilution event cannot occur without reaching the

high level alarm and notify operators.

Calculation No. Fll-0001Attachment 1


2 There should be an acceptance criteriasection which explains why 571 ppm wasused as the final concentration (Cf).Suggestion for wording:

"The worse case misloaded assemblyrequired boron concentration is greater than571 ppmB (Pool B) to maintain a keff lessthan 0. 95. Since it is a higher boronconcentration than that required for Pool A,use of 571 ppmB will result in the leastamount of time from event initiation to thelowest acceptable final boron concentration(Cf). Per IOCFR 50.68(b) (4), The kef for afully unborated condition must still be lessthan 1.0. "

I don't think there is an acceptance criteria forthis calc. The 571 ppm is an input value to thecalculation and produces our results. Making

571 an acceptance criteria makes it seem to melike that is the number we are trying to calculate.

I would think if anything it would be anassumption, however I believe the calc is

sufficient as is. If you disagree we can discussthis further.

3 Under "Purpose", add in that the licensing Donebasis for ITS 4.3.1 is changing for the EPUto credit borated water.

4 Section 2.3. I do not believe that the low Removed reference to ITS.level of 156' 6" bounds the ITS limit (156') inthat use of the ITS value would result in lesstime for operator actions. Suggest removingthe sentence about the ITS.

5 Need to have the Holtec report as a Donereference.

6 Is there any criterion on timing to identify I have discussed time requirements with Opsand terminate a dilution event? The final and we will be implementing the necessaryparagraph states "...there will be sufficient actions in the site APs. The necessary changestime for plant personnel to indentify and will be made to plant procedures as directed by,terminate..." How do we know this? Without EC 71193. I will change it to say "... plantthis information, one can postulate that it will personnel will have 55 minutes to identify andtake the plant longer than 55 minutes to terminate..." so that it does not appear we arereact to a SFP level alarm. saying there is sufficient time without a basis.

7 Please add references for 10x10, 3x3 and References 7.2.4 and 7.2.5 were added.4x3 volume adders.

8 Section 5. Please provide details as to how Changed wording to "... 10 to the estimatedthe bounding analysis is conservative by a worst case realistic SFP dilution flow rate offactor of 10 to the worst case realistic approximately 300 GPM.condition.

Calculation No. F1l1-0001Attachment I


9 Not a deficiency. Independently performed avolume calculation (alternate calculation)and determined that the volumes are asfollows:Pool A - 23558 ft3Pool B - 19003 ft3Total - 42561 ft3The values calculated by in this calculationare smaller and are therefore conservativeand appropriate. The differences areattributable to conservatisms introducedignoring different volumes of water. Notethat the raw volume calculations havealready been design verified and my checkwas to confirm their conservatism.

None required.

10 Not a deficiency. Performed an alternate None required.calculation for the time to reach minimumallowable boron concentrations. Confirmedcalculation of time to alarm.

Crystal River Unit 3 - Response to Request for Additional ...

Documents

Transcript of Crystal River Unit 3 - Response to Request for Additional ...