Letter Sequence Other |
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TAC:45853, (Approved, Closed) TAC:45854, (Approved, Closed) TAC:48192, (Approved, Closed) TAC:48193, (Approved, Closed) TAC:62166, (Approved, Closed) |
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MONTHYEARML20115D7571985-04-0404 April 1985 Withdraws 850402 Exemption Request Re 10CFR50.46 & Implementation of Voluntary Peaking Penalty,Per 850403 Discussion.Penalizing Exxon Fuel by 0.5 Will Conservatively Resolve NRC Concerns Project stage: Other ML20129F0331985-05-24024 May 1985 Forwards Amends 89 & 79 to Licenses DPR-39 & DPR-48, Respectively & Safety Evaluation.Amends Revise Tech Specs Re Limits for Total Nuclear Peaking Factor.Revised Peaking Factor Is 2.32 Project stage: Approval ML20128L6511985-06-26026 June 1985 Discusses Resolution of TMI Action Items II.K.3.30 & II.K.3.31,based on Util Use of Westinghouse Small Break LOCA Model,Notrump.Util Action on Item II.K.3.30 Complete.Generic Ltr 83-35 Requirements for Item II.K.3.31 Provided Project stage: Other ML20134L2351985-08-26026 August 1985 Confirms That Util Is Member of Westinghouse Owners Group & That Notrump Code Will Be Used in Small Break LOCA Analysis for Facility,Per NRC .Tmi Action Item II.K.3.30 Considered Complete.Analysis to Be Submitted Before 850626 Project stage: Other IR 05000266/19860061986-06-11011 June 1986 Insp Repts 50-266/86-06 & 50-301/86-06 on 860401-0531.No Noncompliance or Deviations Noted.Major Areas Inspected: Followup on Items of Noncompliance & Deviations & Operational Safety Verification & ESF Sys Walkdown Project stage: Request ML20206U0381986-07-0101 July 1986 Discusses WCAP-11145,in Response to 850626 Request for Small Break LOCA Analysis.Analysis Unnecessary as WCAP-11145 Complies W/Tmi Action Plan Item II.K.3.31 on Generic Basis Per Generic Ltr 83-35 Project stage: Request ML20206T8351986-07-0101 July 1986 Remits Fee for 860611 Application for Exemptions from Automatic Fire Suppression Sys Requirements Prescribed in Section III.G.3.b of 10CFR50,App R for Component Cooling Water HX & Boric Acid Tank Room Project stage: Request ML20206Q0601986-08-25025 August 1986 Topical Rept Evaluation of Addendum 3 to Rev 1 to WCAP 9561, Thimble Modeling in Westinghouse ECCS Evaluation Model. Changes Meet Requirements of 10CFR50.46 & 10CFR50,App K & Acceptable Project stage: Other ML20214Q3431986-11-20020 November 1986 Forwards Supplemental Safety Evaluation Accepting Util 860611 Response to TMI Action Item II.K.3.31 Re Small Break Loca.Use of Westinghouse Generic Notrump Analysis (WCAP- 1145),approved by NRC on 861006,acceptable.Item Resolved Project stage: Approval 1985-08-26
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Category:TEXT-SAFETY REPORT
MONTHYEARNPL-99-0569, Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with ML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 NPL-99-0051, Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with NPL-99-0449, Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20209D2691999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbnps,Units 1 & 2 ML20196F3341999-06-22022 June 1999 Safety Evaluation for Implementation of 422V+ Fuel Assemblies at Pbnp Units 1 & 2 ML20195F9781999-06-10010 June 1999 Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1 ML20209D2751999-05-31031 May 1999 Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0328, Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with NPL-99-0273, Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With ML20196F3521999-04-30030 April 1999 Non-proprietary WCAP-14788, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power) NPL-99-0193, Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with NPL-99-0134, Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0008, Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with NPL-99-0091, 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with1998-12-31031 December 1998 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 NPL-98-1006, Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20195J5101998-11-16016 November 1998 Proposed Revs to Section 1.3 of FSAR for Pbnp QA Program ML20198J5941998-11-0303 November 1998 1998 Graded Exercise,Conducted on 981103 NPL-98-0948, Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With NPL-98-0880, Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored1998-10-21021 October 1998 Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored ML20154M9121998-10-14014 October 1998 Unit 1 Refueling 24 Repair/Replacement Summary Rept for Form NIS-2 ML20154L6751998-10-14014 October 1998 Unit 1 Refueling 24 ISI Summary Rept for Form NIS-1 NPL-98-0826, Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20151W3851998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Pbnp Units 1 & 2 NPL-98-0653, Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4471998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 2 ML20151W4541998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 1 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 NPL-98-0558, Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 2 ML20151W4261998-06-30030 June 1998 Corrected Page to MOR for June 1998 for Pbnp Unit 2 ML20151W4221998-05-31031 May 1998 Corrected Page to MOR for May 1998 for Pbnp Unit 2 NPL-98-0481, Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4011998-04-30030 April 1998 Corrected Page to MOR for April 1998 for Pbnp Unit 2 NPL-98-0356, Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20216D7071998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3981998-03-31031 March 1998 Corrected Page to MOR for March for Pbnp Unit 2 NPL-98-0209, Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable1998-03-30030 March 1998 Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant NPL-98-0159, Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3891998-02-28028 February 1998 Corrected Page to MOR for Feb 1998 for Pbnp Unit 2 ML20216D7121998-02-28028 February 1998 Revised Corrected MOR for Feb 1998 for Point Beach Nuclear Plant,Unit 2 NPL-98-0084, Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 21998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 1999-09-30
[Table view] Category:TOPICAL REPORT EVALUATION
MONTHYEARML20206Q0601986-08-25025 August 1986 Topical Rept Evaluation of Addendum 3 to Rev 1 to WCAP 9561, Thimble Modeling in Westinghouse ECCS Evaluation Model. Changes Meet Requirements of 10CFR50.46 & 10CFR50,App K & Acceptable 1986-08-25
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o SAFETY EVALUATION ON CHANGES IN THE 1981 WESTINGHOUSE ECCS EVALUATION MODEL WITH BART Introduction I In a meeting held on June 23, 1986, Westinghouse met with representatives of the staff to discuss changes in their large break ECCS Evaluation Models.
These changes were necessitated by a modeling; change in the WREFLOOD code and an input error to the BART code. Westinghouse estimated that inclusion of the modeling change and correction of the input error could result in an increased !
peak clad temperature of up to 120*F for ECCS analyses that were performed with the 1981 Evaluation Model with BART. The corrections and remedial actions are described in Topical Report WCAP 9561 Addendum 3, Revision 1. " Thimble Modeling in Westinghouse ECCS Evaluation Model," which was submitted in a letter dated July 24, 1986 (Reference 1).
The model change in the WREFLOOD code was required because the water volume which flows into the control rod guide thimbles during the core reflooding period following a large break LOCA was previously neglected. This model change was found to have the effect of increasing the calculated peak cladding
- temperature by up to 20*F. The BART change is an error in input which caused l systematically low values of hot assembly bundle power to be used by the i code. This error was found to have the effect of increasing peak cladding temperature by approximately 100 F. To offset these changes, Westinghouse will include a portion of the heat transfer model calculated by BART in the peak claddirig temperature calculation of the LOCTA code. The combined effect of the thimbles on flooding rate and the error in the hot assembly power was determined to be offset by the benefit of including a portion of the heat transfer model. A net benefit was obtained for plants with high peak cladding temperatures approaching 2200*F and a small penalty was obtained for plants with lower peak cladding temperatures.
8609040001 860825 l
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Thimble Filling This issue affects the following Westinghouse ECCS Evaluation Models: 1) 1978;
- 2) 1981; and 3) 1981 with BART. This safety evaluation applies only to the 1981 model with BART. Westinghouse does not intend to correct the 1978 or 1981 models since they will not be used in the future. The control rod guide thimbles are hollow tubes within the fuel rod bundles which replace 25 fuel rods in 17x17 fuel assemblies and 21 fuel rods in 15x15 fuel assemblies.
Control rods are operated within 1/3 of the thimbles. The thimbles may also contain burnable poison rods or in-core neutron detectors. The thimbles contain small flow holes at the bottom to allow water to escape during control rod insertion.
Durtng a large LOCA, the fluid in the thimbles will flash to steam. During the reflooding period, the flow holes will allow water to reenter. In evaluating core reflooding, thimble refilling was not considered by Westinghouse in the WREFLOOD code. This tends to be nonconservative since water which would otherwise enter the coolant channels w7uld instead flow into the thimbles.
Although nonconservative, this simplifying assumption was considered to be reasonable in view of the relatively restricted flow into the thimbles compared to flow through the core and considering the effect of the thimble hole plugging devices. During analysis of the effect of removal of the thimble plugging devices, it was found that plug clearances were sufficiently large and core flow was sufficiently low during reflood that the assumption of no flow in the thimbles warranted reconsideration.
Subsequent work by Westinghouse assessed the effect of the assumption of flow through the thimbles during the reflood phase - a phase in which core flow rates are significantly lower than during blowdown and during which refill of the thimbles would tend to be at the same rate as that of the core. These i subsequent studies indicated that a modeling assumption of flow in the thimbles during reflood would result in a slightly higher peak clad temperature calculation than would result from the assumption of no flow through the thimbles. For the 1981 Westinghouse ECCS Evaluation Model with BART, thimble
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refilling has now been included in a conservative manner in which all thimbles are assumed to be empty at the beginning of the reflooding period and fill at the same rate as the core. This is accomplished by including the total thimble volume in an existing model of the WREFLOOD code which has been approved by the staff (Reference 2). The effect of this conservative modeling of thimble filling during reflood is small (10 to 20 degrees F) because the more significant phemonena of liquid entrainment in the core and steam binding in the coolant loops would be unchanged.
Hot Assembly Bundle Power The hot assembly bundle power error in BART resulted from a confusion between similar input requirements for BART which is utilized to calculate heat trans-fer coefficients and LOCTA which is utilized to calculate peak cladding temperatures. The LOCTA code evaluates the thermal behavior of a single pin and the fluid conditions in the adjacent hydraulic channel. The hydraulic channel is defined by the hydraulic diameter which is a function of the wetted perimeter. The wetted perimeter includes both heated rod surfaces and unheated thimble surfaces. To account for the effect of the unheated surfaces in computing channel enthalpy rise, the total number of fuel pins plus thimbles is input to LOCTA and utilized to calculate a total surface area and an average heat flux for use in the coolant enthalpy rise calculation.
The BART code evaluates an entire fuel bundle including ,the thimbles. In cal-culating the coolant enthalpy rise the BART code correctly utilizes a heated diamciir in defining the coolant channel adjacent to a fuel rod. Only the fuel
~ rod perimeter is utilized to derive the heated diameter and not the thimble perimeter. The use of an average heat flux for both rods and thimbles is therefore not required in BART. Only the number of rods should be input rather than both rods and thimbles as in LOCTA. The same input was utilized for both codes however. This produced an under-prediction in enthalpy rise in BART and caused an over-prediction of the fuel rod heat transfer coefficient to be cal-culated and transferred to LOCTA. The higher heat transfer coefficient caused LOCTA to calculate a peak cladding temperature that was too low by
approximately 100*F. The error is corrected by inputting the number of fuel rods into BART rather than the number of rods plus thimbles. The staff concludes the correction is acceptable.
Heat transfer from the fuel rod surface to the fluid is a combination of con-vection and radiation. Evaluation of data from the FLECHT reflooding heat transfer experiments has shown that radiation represents a significant fraction of the total heat flux. Radiation heat transfer is a function of the fuel rod surface temperature to the fourth power and increases rapidly at elevated tem-peratures. The radiation models in the current BART code were reviewed and approved by the staff as discussed in Reference 2. In transferring the fuel rod heat transfer coefficients from BART to LOCTA for calculation of peak clad-ding temperature, current Westinghouse ECCS evaluations have not included a portion of the radiation heat transfer coefficient. Westinghouse deleted this heat transfer mode because it was thought to have only a small effect on fuel rod cooling and its deletion would provide additional conservatism in their ECCS Evaluation Model with BART. As discussed in Reference 1, the effect of including this portion of radiation heat transfer compensates for the iden-tified hot assembly bundle power and control rod thimble changes in the 1981 ECCS Evaluation Model with BART. Since the staff previously approved the radiation heat transfer model (Reference 2), incorporation of this portion at this time is acceptable.
Conclusion As stated above, the NRC staff concludes that the changes to the 1981 Westinghouse ECCS Evaluation Model with BART, as described in Reference 1, meet the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 and are, therefore, acceptable.
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References
- 1. Westinghouse letter, NS-NRC-86-3147, from E. P. Rahe to J. Lyons, NRC,
" Review of WCAP-9561-P, Addendum 3, Revision 1," July 24,1986.
A
- 2. .NRC letter, from C. O. Thomas to E. P. Rahe, Westinghouse Electric Cor-poration, " Acceptance for Referencing of Licensing Topical Report WCAP-9561, BART A-1: A Computer Code for Best Estimate Analysis of Reflood Transients," December 21, 1983.
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