ML17332A477
| ML17332A477 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/16/1994 |
| From: | Fitzpatrick E INDIANA MICHIGAN POWER CO. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9412230130 | |
| Download: ML17332A477 (23) | |
Text
PRIDRIT'Y Z REGULATOl .IMKRQFIM dI%VIi!dTIOIOYSTEM (RIDE)
ACCESSION NBR:9412230130 DOC.DATE: 94/12/16 NOTARIZED: NO . DOCKET FACIL:50-315- Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co.
RECIP.NAME RECIPIENT AFFILIATION RUSSELL,W.T. Document Control Branch (Document Control Desk)
SUBJECT:
Provides notification of SBLOCA model changes or errors reported by Westinghouse that meet definition of significant as defined in 10CFR50.46.
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PLEASE HELP US TO REDUCE 'O'ASTE! COiNTACTTHE DOCL'Ii!EY'I COYTROL DESK. ROOiiI PI -37 (EXT. SN-2033 ) TO ELI MIYATEYOUR YAME FRO!i!
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Indiana Michigan Power Company P.O. Box 16631 Cofumbus, OH 43216 FI AEP:NRC:1118H 10 CFR 50.46(a)(3)(ii)
Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 REPORT OF LOCA EVALUATION MODEL CHANGES PURSUANT TO 10 CFR 50.46(a)(3)(ii)
U. ST Nuclear Regulatory Commission Document Control Desk Washington, DE C. 20555 Attn: W. T. Russell December 16, 1994
Dear Mr. Russell:
Pursuant to the requirements of 10CFR50.46(a)(3)(ii), this letter provides notification of small break loss of coolant accident (SBLOCA) model changes or errors reported to us by Westinghouse Electric Corporation (Westinghouse) that meet the definition of significant as defined in 10CFR50.46.
Attachment 1, which was provided to us by Westinghouse, describes errors discovered in their NOTRUMP computer code and changes to the code. NOTRUMP is used for small break LOCA analysis for Units 1 and 2 of Donald C. Cook Nuclear Plant. The code does not address ECCS switchover to recirculation which results from rapid draining of the refueling water storage tank due to automatic actuation of the containment spray system. Westinghouse concludes that this condition necessitates a peak clad temperature (PCT) penalty of 20'F for some plants. For Cook Nuclear Plant, this penalty is only applied to the high head safety injection (HHSI) Cross-Tie Valve Open case at 3588 MWt for both Unit 1 and Unit 2.
Attachment 2, which was provided by Westinghouse, also describes changes and errors in the NOTRUMP computer code. The errors involve the boiling heat transfer correlation, steam line isolation logic, and initialization of the cladding zirconium oxide thickness prior to creation of fuel zones analogous to the mixture and vapor regions for core nodes.
9412230130 941216T 1 PDR ADOCK 05000315 P PDRQ 3 7
Mr. W. T. Russell AEP'NRC'1118H The requirement for reporting LOCA model changes/errors under 10CFR50.46 are not applicable until the sum of the absolute values of the PCT penalties accumulate to a change greater than 50 F. Since none of the LOCA cases had a PCT summation greater than 50'F, a 10CFR50.46 report was not filed at the time of receipt of either Attachment 1 or Attachment 2. , which was provided to us by Westinghouse, also describes changes and errors in small break LOCA codes. The issue concerns a deficiency in the amount of detail used for the axial nodalization of the fuel rod, as it affected the solution of the channel fluid equations. Further investigation by Westinghouse identified several additional related issues, as described in Attachment 3.
The absolute value of the sum of the changes and errors described in Attachments 1, 2, and 3 indicate PCT changes greater than 50'F, Since the changes in PCT are more than 50'F, the changes meet the definition of significant provided in 10CFR50.46. This report is being provided in response to the guidance in 10CFR50.46.
During the review of the peak clad temperatures displayed in , it was discovered that the Unit 1 SBLOCA PCT margin calculation for the 3250 MWt case with the HHSI cross-tie valves closed contains an assessment which should have been removed following the reanalysis of this case in December 1993. The assessment of -13~F for drift flux flow regime errors should no longer be tracked because the version of the NOTRUMP code used for that reanalysis includes the corrected drift flux flow regime map. Further discussion of this item is provided in the Westinghouse letter, which is given as Attachment 4.
The corrected Attachment 5 contains the peak clad temperatures calculated specifically for Donald C. Cook Nuclear Plant Units 1 and 2. In all cases, the calculated peak clad temperatures remain within the 10 CFR 50.46 limit of 2200 F.
This 10 CFR 50.46 report has been prepared based on analyses performed for the MSSV setpoint tolerance relaxation. These analyses were recently reviewed and approved in conjunction with Amendment 182 to Facility Operating License DPR-58 and Amendment 167 to Facility Operating License DPR-74.
The plan for revising the Unit 1 cross-ties open SBLOCA run, discussed in our letter designated AEP:NRC:1118G, is being delayed several months to ensure that the calculation is made
Mr. W. T. Russell AEP:NRC:1118H with planned code revisions. Current plans also include both LBLOCA and SBLOCA'reanalyses in conjunction with evaluations and analyses to support an increase in allowable steam generator tube plugging (SGTP) for Unit 1. This work is tentatively planned to be complete and submitted to the staff by June 1995. Plans for new LBLOCA and SBLOCA analyses of record for Unit 2 remain tentative at this time.
Sincerely,
~ Vice E. E. Fitzpatrick
~ ~
President sic Attachments cc: A. A. Blind - Bridgman G. Charnoff J ~ B. Martin - Region III NFEM Section Chief NRC Resident Inspector - Bridgman J. R. Padgett
ATTACHMENT 1 TO AEP:NRC:1118H WESTINGHOUSE ELECTRIC CORPORATION DESCRIPTION OF LOCA MODEL CHANGES
t 4 ~
RscRm A potential issue has been identified related to automatic actuation of the containment spray system during certain small break LOCAs. Through analyses it has been found that breaks as small as taro inches or less may actuate the containment spray system within several minutes of break initiation for some plants.
As a result of the high containment spray flow rate and the prolonged nature of the small break LOCA, switchover to sump recirculation may be required before the event is completely resolved (ECCS injection flow exceeds break fiow, RCS mass is increasing and peak clad temperature is decreasing or the core is fully quenched). The current small break licensing basis does not address several issues related to ECCS switchover and long term recirculation.
eD ti n Section D of 10 CFR 50, Appendix K requires that the containment pressure used in the ECCS evaluation models.3ccount for the effects of all installed containment heat removal equipment. The current Westinghouse small break ECCS evaluation model is riot considered to be sensitive to containment pressure effects because the calculated break flow is critically limited throughout the transient. As such, the small break ECCS evaluation model does not account for actuation and operation of the containment spray system. Furthermore, since the small break ECCS evaluation model does not account for operation of the containment spray system, the ECCS evaluation model does not include switchover to ECCS recirculation. This approach may no longer be bounding with respect to maximum PCI'.
The current Westinghouse small break evaluation model assumes a continuous supply of ECCS water at injection mode flow rates and the enthalpy of the injection Quid throughout the transient. At the time of switchover to recirculation during a small break LOCA, the ECCS pumps may be shut down for realignment resulting in a period of no SI delivery for some plants. In addition, some plants may have less available ECCS recirculation Qow than ECCS injection flow due to closure of the safety injection pump discharge cross-tie valves. Furthermore, the current small break licensing basis analysis does not account for the change in enthalpy due to recirculation since it had been assumed that the event would be resolved before recirculation.
This issue also has the potential to effect the limiting single failure assumption used in the small break ECCS evaluation model. The current single Mute assumption is the Mure of an emergency diesel generator and loss of one entire safeguards train. Ifboth emergency diesel generators start and supply power to the containment spray pumps, the RWST would drain earlier in the transient. The loss of a single intermediate head pump or charging pump could conceivably be more limiting than the loss of a complete train of safeguards systems.
The modeling of these effects could potentially lead to worse results than those predicted in the current analyses, challenging the 10 CFR 50.46 acceptance criteria.
HSAL94410
T cal v uatio In support of this issue, a number of plant specific and bounding generic evaluations were performed.
First, all plants for which Westinghouse performs the small break ECCS analysis and known to have either interrupted or reduced ECCS flow as a result of switchover to ECCS recirculation were identified:
D. C. Cook 1 &2 Point Beach 1 & 2 Ginna Prairie Island 1 & 2 Kewaunee Turkey Point 3 & 4 Kori 2 For these plants, it was shown that even with an alternate single failure which would more rapidly drain the RWST, the current small break analyses of record would remain the most limiting analysis with respect to maximum calculated PCI'. Thus, for these plants with interrupted or reduced ECCS Qow as a result of switchover to ECCS recirculation, the acceptance criteria of 10 CFR SOA6 is met and does not represent any defect, deviation or failure to comply with respect to 10 CFR 21. Furthermore, no PCT penalty or benefit assessment is required as a result of interrupted or reduced ECCS Qow.
Second, it was determined that all plants currently licensed with the Westinghouse small break ECCS evaluation model were potentially impacted by the inc;ease in ECCS water enthalpy when the ECCS is switched over from the injection mode to the recirculation mode of operation. Through the use of engineering analysis including an alternate single failure which would more rapidly drain the RWST, it was determined that a number of, plants were not affected by this issue in terms of peak clad temperature (PCT). However', a limited number of plants are impacted by the increase in ECCS water enthalpy during recirculation:
Beaver Valley 1 & 2 Shearon Harris D. C. Cook 1 & 2 Yonggwang 2 Indian Point 2 Surry 1 & 2 Indian Point 3 V. C. Summer J. M. Farley 1 & 2 Kori3 &4 North Anna 1 & 2 For these plants it was determined that a 20'F PCI'enalty assessment is sufficient to account for the increase in ECCS water enthalphy during recirculation. For these plants only, plant specific Small Break PCT Margin Utilization Summary sheets are provided.
For those plants for which Westinghouse performs the licensed SBLOCA analyses, Westinghouse has determined that this issue is not a substantial safety hazard pursuant to 10 CFR 21 because the PCT penalty does not result in a loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety. However, for those plants assessed a 20'F PCI'enalty, the plant licensees should review their reporting obligations under 10 CFR 50.46. To facilitate this, the updated Small Break PCT Margin Utilization Summary sheets are attached to this letter for those plants.
ATTACHMENT 2 TO AEP:NRC:1118H WESTINGHOUSE ELECTRIC CORPORATION
'~q.,-iDESCRIPTION OP LOCA MODEL CHANGES
BOILING HEAT TRANSFER CORRELATION ERRORS Qggkkgggg This closely related set of errors deals with how the mbrture velocity is defined for use in various boiling heat transfer regime correlations. The previous definition for mixnue velocity did not properly account for dry and slip effects calculated in NOTRUMP. Ignis error particularly af5xted NOTRUMP calculations of heat transfer coefficient when using the Westinghouse Transition Boiling Correlation and the Dougall-Rohsenow Santrated Film Boiling Correlation.
In addition, a minor typographical error was also corrected in the Westinghouse Transition BoBing Correlation.
This was determined to-be a No'n-Discretionary Change as described in Section 4.12 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
ected valu ti n M de 1985 Small Break LOCA Evaluation Model Representative plant calculations for this issue resulted in the estimated PCT effect documented in the attached Margin Utilization Sheet.
SXKQN LINE ISOLATXON LOGIC ERRORS This error consists of two portions: a possible plant specific effect which only applies to analyses which assumed Main Feedwater Isolation (FWQ to occur on Swignal, and a generic effect applying to all previous analyses.
The possible plant specific effect was the result of incorrect logic which caused the main steam line isolation to occur on the same signal as FWI. 'Iherefore, when the S-signal was chosen through user input to be the appropriate signal for FWI, it also caused the steam line isolation to occur on S-signal.
This is inconsistent with the standard conservative assumption of steam line isolation on Loss of Offsite Power coincident with the earlier Reactor Trip signal.
The generic effect was the result of incorrect logic which always led to the isolation functions occumng at a slightly later time than when the appropriate signal was generated.
This was determined to be a Non-Discretionary Change as descnbed in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
v a 1985 Small Break LOCA Evaluation Model a
Representative plant calculations for this issue resulted in the estimated PCI'ffect (+ 12'F for the plant specific portion, if applicable, and +18'F for the generic portion) documented m the attached Margin Utilization Sheet.
CORE NODE ZIRC OXIDE NlTJ4URA D 0 NOTRUMP models two regions for each core node analogous to the two (mixture and vapor) regions in adjoining fluid nodes. During the course of a transient, NOTRUMP tracks region specific quantities for each core node. Erroneous logic caused incorrect initialization of the region speci6c, fuel cladding zirc oxide thickness at times prior to the actual creation of the relevant region during the core boilifftransient.
This was deteanined to be a Non-DiscretionarJJ Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
valua de M n ~
1985 Small Break LOCA Evaluation Model Representative plant calculations led to an estimated generic PCT effect of O'F for this effect.
~ ~
ATTACHMENT 3 TO AEP:NRC: 1118H WESTINGHOUSE ELECTRIC CORPORATION DESCRIPTION OF LOCA MODEL CHANGES
~gkg~und 10CFR50.46, Appendix K prescribes the acceptable features and required documemation for ECCS Evaluation Models. More specifically,Section II.3 requires that documentation be in place to verHy that sensitivity studies have demonstrated the adequacy of nodalization schemes used in the analysis models. A study.was recently underuiken with the Westinghouse small break LOCA Evaluation Model to examine the sensitivity of predicted results to the nodalizatioa used for ihe hot rod model.
'Ihe results of that study raised concerns regarding the adequacy of the standard axial nodalizatioa prescribed for use in the SBLOCI'A code for licensing basis analyses. As a amdt of this concern, Westinghouse iavestigated this as a Potemial Issue per 10 CFR 21.
'Ihe standard rod model. (developed in the 1970's) used in performing SBLOCTA calculations has 19 axial nodes with a finer distribution in the top elevations. However, sensitivity studies to justify the number and distribution of these aodes can not be documented. A series of calculations were performed using increasingly finer axial nodalizations than prescribed for the 19 node model and indicated that the standard SBLOCTA 19 node model was not conservative. Nearly all cases demonstrated a significantly non~nservative behavior with respect to PCT. The penalty is attributed to a aet increase ia singl~hase steam eathalpy rise as these nodes uncover sooner a'nd heat up more thaa coarser nodes partially covered by the mixture level. Thus, it was concluded that a revised model that included a much finer axial nodalizatioa could potentially lead to less favorable results than those predicted in the current analyses, possibly challenging the 10 CFR 50.46 acceptance criteria.
As a result of further investigation into the SBLOCI'A code, several addiYioaal related issues associated with nodalizatioa and the overall solution of the Quid conservaaon equations were subsequently identified and corrected. As a separate, but related issue, Westinghouse has implemented a revised model for calculating transient fuel rod internal pressure in the SBLOCTA code. Fuel rod pressure is a governiag fhctor in defining the clad creep, burst and blockage behavior for small break LOCA transients. The NRC was informed of this modeling change per Westinghouse letter NTD-NRC-944253, "Revision to the Rod Internal Pressure Model in the Westinghouse SBLOCTA Code (Proprietary)". The letter also informed the NRC that Westinghouse has validated and instituted the model as a methodology improvement to the small break LOCA model for standard
'implementation on a forward-fit basis ia accordance with WCAP-13451, Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting, October, 1992.
caJ valuati At this time Westinghouse has completed the generic technical evaluation of the fuel rod axial nodalization methodology. A revised standard for rod nodalization has been established which insures an adequate solution to the hot channel calculation by specifying a fine nodalization of 0.25 f't nodes for all elevations that are predicted to uncover during the transient.
Since the improved axial aodalization methodology and revised fuel rod internal pressure model can have significant synergistic effects on the predicted peak clad temperature, the SBLOCTA calculation
Rom the limiting small break LOCA transient has been rerun with the revised code and methodology in order to obtain an accttrate estimation of the net effect of these changes on the analysis of tecord.
Several recent code revisions and error corrections of lesser magninide have also been incorporated in the code version used to conduct this calcuhtion. Normally these items would have been teported in the 10CFRSOA6 year~d reporting summary along with estimates of effects. As a consequence of using the revised code to obtain results for this evaluation, these items have been implicitlyaddressed been reanalyzed, Westinghouse believes that no additional reanalysis is necessary to S0.46 even for those plants that have a significant PCI'hange as a result of this issue.
~I in the results provided. Since this portion of the ECCS Small Break Evaluation Model has already CFR Since all of the issues relate to portions of the SBLOCI'A code and/or its associated input methodology, they may be reported as a single closely-related group of changes. Attached to this letter is a revised Small Break LOCA Margin Utilization Summazy table which comains a compilation of the net effect of this evaluation, as item "Axial Nodalizaiion, RIP Model Revision and SBE.OCI'A Error Corrections". Where necessary, Westinghouse has provided notes as an attaclimimt to explain which items have been affected.
P w r n tilizatio During the process of reviewing the analysis of record for D. C. Cook Unit 1 as part of addressing the above issues, conservatism was noted in the core power axial ofBet limit and the hot assembly peaking factors assumed in the Unit 1 analysis. The overall current licensing basis analyses restrict the axial offset to a maximum full power positive skew of 13%, and a maximum hot assembly peaking factor of 1.38. Additional conservatism had been incorporated into the smaH bteak LOCA
. anaiysis to provide margin above and beyond the present core design limits. Following consultation with cognizant core design and utility personal, it was concluded that this margin is not being utQized and could be made available to offset the penalty associated with resolution of the present issues. The
~ . revised calculation. was therefore performed with an axial offset limit of 20% and hot assembly peaking factor of 1.38 which supports the same RSAC core design limits as previously. supported by the analysis of record, and therefore there are no changes to plant Tech Specs Rom incorporating this revision.
R mm ndati For those plants for which Westinghouse performs the licensed SBLOCA analyses, Westinghouse has determined that this issue is not a substantial safety hazard pursuant to 10 CFR 21 because the does not result in a loss of safety Ruction to the extent that there is a major reduction in the PCI'enalty degree of protection provided to public health and safety. However, for those plants that have been assessed either a PCT penalty or benefit, the plant licensees should review their repomng obligations under 10 CFR 50A6. To facilitate this, the updated Small Break PCT Margin Utilization Summary sheets are attached to this letter for those plants.
ATTACHMENT 4 TO AEP'NRC:1118H WESTINGHOUSE ELECTRIC CORPORATION DESCRIPTION OF LOCA MODEL CHANGES
Revised PCT hlargin Utllhation Sheets for D. C. Cook Units 1 and 2 r It has been noted that the D. C. Cook Unit 1 Small Break LOCA PCT Margin Udllzadoa Sheet for the 3250 hGVt case with KHSI cross-tie valves closed contains an assessment which should have been removed following the reanalysis of this case last December (SECL-91<29, Rev2, AEP-93-252). The assessment of -13'F for Drift Hux How Regime Errors should no longer bs tracked because the version of the NOTRUMP code used for thc reanalysis includes the corrected drift flux flow regime map. The corrected current PCT Margin Utilization Sheets are amtclM.d.
Note that the original transmittal of the reanalysis results contained only the analysis PCT (1951'P) and thc corresponding Burst and BlockagefQme ln Life assessment of 117'P. No additional assaements were considered in the Burst and Blockage/Gmc in Life calculatioa 'Ihercfore thc total PCT given in AEP-93-252 was not affected by the Drift Hux Flow Regime Error assessment.
'Ihc Burst and Blockage/Time in Life assessment has been recalculated for all of thc previous PCT Margin
- Utilization Sheets for the reanalysis case (with the -13'F assessmcnt removed). Note that thc Burst and BlockagefTime ia Life assessmeats given here decreased (rather than increased) because a later version of the SPIKE code was used here, incorporating modiflcatioas to the cladding burst strah modeL NSAL-94-022A, October 25, 1994 ansmitted in AEP-94-253, 10/27/94)
Adjusted Burst and BlockagefQme ia.Life assessment: 15'F (No Change)
NSAL-94418A, August 17, 1994 (Thmmittcd ln AEP-.94-247, 8/18/94)
Adjusted Burst and Blockage/Iimc in Life assessment: 90'F NSAL-944lOA, May l6, 1994 (Transmitted la AEP-94-234, 5/17/94)
Adjusted Burst aad BlockagefIime ia Life assessment: 79'F NSAL-94-004A, February 8. 1994 (Transmitted in AEP-94-214, 2/8/94)
Adjusted Burst and BlockagefIlme in Life assessment: 79'F M
At ao time since thc reanalysis has thc total PCT (without the -13'F assessment and with any resulting incremental Burst aad BlockagefHme in Life penalty} exceeded the 2200'F acceptance criterion of 10 CPR 50.46.
ATTACHMENT 5 TO AEP:NRC: 1118H WESTINGHOUSE ELECTRIC CORPORATION DETERMINATION OF EFFECT OF LOCA MODEL CHANGES ON COOK NUCLEAR PLANT LOCA ANALYSES to AEP:NRC:1118H Page 1 SMALL BREAK LOCA PLANT NAME: DONALD C. COOK NUCLEAR PLANT UNIT 1 Comments: gvaluation Model:QOO'~, FQ ~3, F~H ] 55, SGTP +5X Other: HHSI Cross Tie Valve Closed, 3250 MWt Reactor Power A. ANALYSIS OF RECORD PCT ~95 oF B. PRIOR LOCA MODEL ASSESSMENTS - 1992 APCT + 30F~
C. PRIOR LOCA MODEL ASSESSMENTS - March 1994 4PCT -16OF D. 1994 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)
- 1. Boiling Heat Transfer Correlation Error APCT -6 F
- 2. Steam Line Isolation Logic Error 4PCT + 18 F
- 3. Axial Nodalization, RIP Model Revision, and SBLOCTA Error Corrections Analysis APCT -235'Fz
- 4. Burst and Blockage/Time in Life APCT + 150F~
LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT ~130 F The 1992 assessment for 15x15 hydraulic test results was not included in the new analysis of record. However, the drift flux flow regime error was incorporated. The drift flux flow regime assessment was erroneously reported as
-13 F in our previous two reports of January 12, 1994 (AEP:NRC:1118G) and March 25, 1994 (AEP:NRC:1118E). In both cases, the total PCT did ~n t exceed the 22004F acceptance criterion of 10CFR50.46.
- 2. Based on limiting case reanalysis with reduced axial offset (20X) and core radial peaking factor (1.38).
- 3. It should be noted that the burst and blockage assessment is subject to change as other model assessments are made because the magnitude of the burst and blockage assessments depends on the PCT without burst and blockage.
to AEP:NRC:1118H Page 2 SMALL BREAK LOCA PLANT NAME: DONALD C. COOK NUCLEAR PLANT UNIT 1 Comments: Evaluation Model:~OVUM, FQ-~3, FaH ] 55, SGTP QSX Other: HHSI Cross Tie Valve ~e , 3588 MPt Reactor Power A. ANALYSIS OF RECORD PCT ~57C F B. PRIOR LOCA MODEL ASSESSMENTS - October 1993 4PCT -130F C. PRIOR LOCA MODEL ASSESSMENTS - January 1994 4PCT~ + 970F D. PRIOR LOCA MODEL ASSESSMENTS - January 1994 6PCT ~6oF E. 1994 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)
- 1. Containment Spray during SBLOCA dPCT~ + 20OF 2 Boiling Heat Transfer Correlation Error IMP CT -6oF 3
~
~ Steam Line Isolation Logic Error 6PCT ~+8oF
- 4. Axial Nodalization, RIP Model Revision, and SBLOCTA Error Corrections Analysis hPCT -1184Fi F. LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT-~155 'F Based on-limiting case reanalysis with reduced axial offset (20') and core radial peaking factor (1.38).
to AEP:NRC:1118H Page 3 SMALL BREAK LOCA PLANT NAME: DONALD C. COOK NUCLEAR PLANT UNIT 2 Comments: Evaluation Model:QOO~P,, ~g 45~, FwH ] 666, SGTP +5X Other: HHST Cross Tie Valve ~C osed, 3250 HWt Reactor Power A. ANALYSIS OF RECORD PCT 19560F B. PRIOR LOCA MODEL ASSESSMENTS - October 1993 RPCT ~3 P C. PRIOR LOCA MODEL ASSESSMENTS - March 1994 hPCT -160F D. 1994 10CFR50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)
- l. Boiling Heat Transfer Correlation Error hPCT -6OF
- 2. Steam Line Isolation Logic Error hPCT + 18OF
- 3. Axial Nodalization, RIP Model Revision, and SBLOCTA Error Corrections Analysis 6PCT ~+5OoFz Ps
- 4. Burst and Blockage/Time in Life hPCT ~
E. LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT- l9¹6P
- 1. The Fo supported was previously reported incorrectly as 2.357.
- 2. The evaluation case used to determine this assessment also predicted rod burst at the beginning of fuel life.
Therefore, burst and blockage effects are included here and the Burst and Blockage/Time in Life assessment is zero. It is possible for a non-burst case, in combination with possible future permanent model assessments and associated Burst and Blockage/Time in Life penalty, . to become limiting. If such a case becomes limiting, this assessment will change to reflect the non-burst case, and burst and blockage effects will be accounted for below.
- 3. It should be noted that the burst and blockage assessment is subject: to change as other model assessments are made because the magnitude of the burst and blockage assessments depends on the PCT without burst and blockage.
to AEP:NRC:1118H Page 4 SMALL BREAK LOCA PLANT NAME: DONALD C. COOK NUCLEAR PLANT UNIT 2 Comments: Evaluation Hadal:HODEUNP, FQ E 44~, FaH-1 644, SCTP 153 Other: HHST Cross Tle Valve ~C osed, ~34 3 HWt Reactor Power A. ANALYSIS OF RECORD PCT ~947'F B. PRIOR LOCA MODEL ASSESSMENTS - October 1993 dPCT~ -130F C. PRIOR LOCA MODEL ASSESSMENTS - March 1994 dPCT-~6 F D. 1993 10CFR50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)
- 1. Boiling Heat Transfer Correlation Error APCT -6oF
- 2. Steam Line Isolation Logic Error hPCT + 18~F
- 3. Axial Nodalization, RIP Model Revision, and SBLOCTA Error Corrections Analysis hPCT -45 F Burst and Blockage/Time in Life 4PCT + 58 Fz E. LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT ~943oF
'4 ~
The supported was previously reported incorrectly as 2
- 2. It should be noted that the burst and blockage assessment is subject to change as other model assessments are made because the magnitude of the burst and blockage assessments depends on the PCT without burst and blockage.
to AEP:NRC:1118H Page 5 SMALL BREAK LOCA PLANT NAME: DONALD C. COOK NUCLEAR PLANT UNIT 2 Comments: Evaluation Model:NOTRUMP, ~~3 , FaH ~6 ,SGTP 15X, Other: HHSI Cross Tie Valve ~e , ~3 88 MVt Reactor Power A. ANALYSIS OF RECORD PCT~ 1531~F B. PRIOR LOCA MODEL ASSESSMENTS - October 1993 IKPCT 13oF C. PRIOR LOCA MODEL ASSESSMENTS - March 1994 APCT -16oF 1993 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)
- l. Containment Spray During SBLOCA APCT + 20'F
- 2. Boiling Heat Transfer Correlation Error EPCT -60F
- 3. Steam Line Isolation Logic Error hPCT + 18oF