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| {{#Wiki_filter:CENPD-279 SUPPLEMENT 1 ANNUAL REPORT ON'C-E ECCS CODES AND METHODS FOR 10CFR50.46 FEBRUARY.1990 9003260577 900228 PDR AQOCK 05000~28 P INelggp~gg | | {{#Wiki_filter:CENPD - 279 SUPPLEMENT 1 ANNUAL REPORT ON |
| ~I i+[C'UP LEGAL NOTICE This report was prepared as an account of, work sponsored.by Combustion Engineering, Inc.'Neither.Combustion Engineering nor any person acting on its.behalf: A.makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose.or merchantabil,ity, wi.th respect to the accuracy, completeness, or usefulness of the information contained in this report, or'that the use-of any information, apparatus, method, or process disclosed''in.this report may not infringe privately owned rights;or B.assumes any l,i'abilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed'in this report.
| | 'C-E ECCS CODES AND METHODS FOR 10CFR50.46 FEBRUARY. 1990 9003260577 900228 PDR AQOCK 05000~28 P |
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| CENPD-279 SUPPLEMENT, 1 ANNUAL REPORT ON C-E ECCS CODES AND METHODS FOR 10CFR50.46 TRANslENT METHoos ANo LOCA NUCLEAR FUEL ENGlNEERING FEBRUARY 1990 41 4>
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| Abstract This report describes changes and errors in the Combustion Engineering codes and anal'ysis methodology for ECCS analysis in 1989 per the requirements of 10CFR50.46.
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| For this reporting period only one computer code had reportable changes or errors.The corrections and changes did not affect the peak cladding temperature.
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| The,cummulative temperature change for large break LOCA is a reduction of less than.1 F.No changes or errors..that affect the peak cladding temperature for small break LOCA have occured.Per the:.cri.teria of 10CFR50.46, no action.beyond this annual report is required.
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| 4i i+i 4 ll Table of Contents 1.0'Introduction 2.0 Codes for ECCS Evaluation 3.0 Error Corrections.
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| and Model Changes in Computer Codes 3.1 COMPERC-II 4',0 Conclusions 5.0 References 0 0
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| ==1.0 Introduction==
| | LEGAL NOTICE This report was prepared as an account of, work sponsored .by Combustion Engineering, Inc. 'Neither. Combustion Engineering nor any person acting on its .behalf: |
| This report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis.The revision to the ECCS Acceptance Criteria spells out reporting requirements and actions required when errors are corrected or changes are made in an evaluation model or in the application of a model for an operating licensee or construction permittee of a nuclear power plant.The action requirements in g 50.46(a)(3) are: l.Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. | | A. makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose. or merchantabil,ity, wi.th respect to the accuracy, completeness, or usefulness of the information contained in this report, or 'that the use- of any information, apparatus, method, or process disclosed''in |
| For, this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature (PCT)different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 F.2.For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in g 50.4..This report is,to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a revised evaluation correcting minor errors is approved by the NRC staff.3.If the change or error is significant, the applicant or licensee shall provide this report.within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with g 50.46 igi'I I requirements.
| | .this report may not infringe privately owned rights; or B. assumes any l,i'abilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed 'in this report. |
| This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC.For those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the proposed schedule.4.Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in paragraph (b)of g 50.46 is a reportable event as described in gg 50.55(e), 50.72 and 50.73.The affected applicant or licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with g 50.46 requirements. | | |
| This report documents all the changes made to the presently licensed C-E LOCA analysis models and methodology which have not been reviewed by the NRC staff.This is specifically to satisfy the requirements described in the second item above.2.0 Codes for ECCS Evaluation C-E uses several digital computer codes for ECCS analysis that are described in topical reports, are licensed by the NRC and are covered by the provisions of CFR 50.46.Those for large break LOCA.calculations are: CEFLASH-4A, COMPERC-II, PARCH, STRIKIN-II, and COMZIRC.CEFLASH-4AS is used in conjunction with COMPERC-II, STRIKIN-II, and PARCH for small break LOCA calculations.
| | CENPD - 279 SUPPLEMENT, 1 ANNUAL REPORT ON C-E ECCS CODES AND METHODS FOR 10CFR50.46 TRANslENT METHoos ANo LOCA NUCLEAR FUEL ENGlNEERING FEBRUARY 1990 |
| 3.0 Error Corrections and Model Changes in Computer Codes\This section discusses all error corrrections or model changes to the licensed codes which may affect the calculated PCT.Only the COMPERC-II for a large break has been changed in 1989.No changes to analysis procedures have been made since the, last approved submittal to the NRC. | | |
| ~l 3..I., COMPERC-I I A.Code Description COHPERC-II is a FORTRAN digital computer program which is used by Combustion Engineering, Inc.to calculate the core refill and reflood transient portion of a PWR loss of coolant accident (LOCA).A detailed code.description is presented in References 2 through 4.B.Model Change in COMPERC-II for SI.Spillage The model for the spil.lage calculation.in COMPERC-II'has been changed.to reflect a more realistic physical representation.
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| The change.is as described below: Nd1 (Pg 10 f'Rf 2)'f ZA, HAX ZA-ZA, MAXI.2 A 2 I/2 A'MAXI A MAX c'B F spill spill where Wspill spill B F ZA A,MAX'A, MAXI'A gc Rate of water spillage out of the break, Loss coefficient for, the spillage of water out of the break,.Flow, area in the core, Height of the water in the downcomer, Distance between bottom of core and.bottom of inlet pipe, Dis'tance between bottom of core and top of inlet pipe, Density of water in the downcomer/lower plenum, Conversi'on constant.3 ii*I Modification Equation (I)was modified as (1A spill 2 2 I/2 A MAX c A B F spill (2)The difference between Equations (I)and (2)is the first term of the numerator on the right hand side of the equations.
| | Abstract This report describes changes and errors in the Combustion Engineering codes and anal'ysis methodology for ECCS analysis in 1989 per the requirements of 10CFR50.46. For this reporting period only one computer code had reportable changes or errors. The corrections and changes did not affect the peak cladding temperature. The,cummulative temperature change for large break LOCA is a reduction of less than. 1 F. No changes or errors..that affect the peak cladding temperature for small break LOCA have occured. Per the:.cri.teria of 10CFR50.46, no action .beyond this annual report is required. |
| This change uses the real head term instead of the fixed head term while the mixture level is in the span of the cold.leg.C.Reasons for the Modification As indicated in the previous section, this change reflects a better physical representation than the model described in Reference l.However, a more important reason for this modification is to remove low-amplitude flow oscillations introduced by the discontinuity of the fixed head in the old model.D.Impact of the Spil.lage Model Change on PCT The change in downcomer spillage head term has the possibility to affect PCT through two effects--reflood rate and two-phase level.Comparison of the reflood rates for cases without and with the change in head term shows that the small oscillations in the reflood rate are removed.However, the reflood rate selected for subsequent use is not changed;therefore, there is no change in PCT from this effect.The change in the head term for downcomer spillage also eliminates oscillations in the two phase-level but does not change the base two-phase level.Elimination of the oscillations reduces the uncertainty in the two-phase level selected for the next step in the analysis.However, due: to the small'ensitivity of the C-E methodology to two-phase level changes,.the change in the two-phase il ll,!I level due to the change in the head term for spillage has no effect on PCT.4.0 Conclusions The change to COMPERC-II has the potential to affect the PCT by changing the reflood rate or the two-phase level.However, an evaluation, of the reflood rates and effect of the two-phase level for cases before and after the change in head term for the downcomer spillage shows that there is no change in PCT.The cummulative change in.PCT for large break LOCA including that from the previous annual report, Reference 5, is a reduction of, less than 1 F.There have been no changes in'he small break LOCA results to date.Therefore, there was no significant change in the sense of CFR 50.46 in 1989 and no action beyond the submission of this report is needed.5.0 References 1."Emergency Core Cooling System;Revisions to Acceptance Criteria," 10CFR50, Federal Register,, Vol.53, No.180, September 16, 1988.2.CENPD-134P,"COMPERC-II, A Program for Emergency-Refil.l-Reflood of the Core," August, 1974.3.CENPD-134P, Supplement 1,."COMPERC-II, A Program for Emergency'Refill-Reflood of the Core (Modifications)," February, 1975.4.CENPD-134, Supplement 2,"COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June, 1985.5.CENPD-279, Annual Report on C-E.ECCS Codes and Methods for 10CFR50.46, April, 1989., 5 4l i+i a>>Z}}
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| | Table of Contents 1.0 'Introduction 2.0 Codes for ECCS Evaluation 3.0 Error Corrections. and Model Changes in Computer Codes |
| | : 3. 1 COMPERC-II 4',0 Conclusions 5.0 References |
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| | 0 0 1.0 Introduction This report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis. The revision to the ECCS Acceptance Criteria spells out reporting requirements and actions required when errors are corrected or changes are made in an evaluation model or in the application of a model for an operating licensee or construction permittee of a nuclear power plant. |
| | The action requirements in g 50.46(a)(3) are: |
| | : l. Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For, this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature (PCT) different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 F. |
| | : 2. For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in g 50.4.. This report is,to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a revised evaluation correcting minor errors is approved by the NRC staff. |
| | : 3. If the change or error is significant, the applicant or licensee shall provide this report. within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with g 50.46 |
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| | requirements. This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC. |
| | For those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the proposed schedule. |
| | : 4. Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in paragraph (b) of g 50.46 is a reportable event as described in gg 50.55(e), 50.72 and 50.73. The affected applicant or licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with g 50.46 requirements. |
| | This report documents all the changes made to the presently licensed C-E LOCA analysis models and methodology which have not been reviewed by the NRC staff. This is specifically to satisfy the requirements described in the second item above. |
| | 2.0 Codes for ECCS Evaluation C-E uses several digital computer codes for ECCS analysis that are described in topical reports, are licensed by the NRC and are covered by the provisions of CFR 50.46. Those for large break LOCA .calculations are: CEFLASH-4A, COMPERC-II, PARCH, STRIKIN-II, and COMZIRC. CEFLASH-4AS is used in conjunction with COMPERC-II, STRIKIN-II, and PARCH for small break LOCA calculations. |
| | 3.0 Error Corrections and Model Changes in Computer Codes |
| | \ |
| | This section discusses all error corrrections or model changes to the licensed codes which may affect the calculated PCT. Only the COMPERC-II for a large break has been changed in 1989. No changes to analysis procedures have been made since the, last approved submittal to the NRC. |
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| | ~l 3 ..I., COMPERC- II A. Code Description COHPERC-II is a FORTRAN digital computer program which is used by Combustion Engineering, Inc. to calculate the core refill and reflood transient portion of a PWR loss of coolant accident (LOCA). |
| | A detailed code. description is presented in References 2 through 4. |
| | B. Model Change in COMPERC-II for SI. Spillage The model for the spil.lage calculation .in COMPERC-II'has been changed .to reflect a more realistic physical representation. The change .is as described below: |
| | Nd1 (Pg 10 f'Rf 2)'f ZA, HAX ZA ZA, MAXI c' |
| | 2 A |
| | 2 I/2 A'MAXI A MAX B F spill spill where Wspill Rate of water spillage out of the break, spill Loss coefficient for,the spillage of water out of the break,. |
| | B F Flow, area in the core, ZA Height of the water in the downcomer, A,MAX Distance between bottom of core and .bottom of inlet pipe, Dis'tance between bottom of core and top of inlet |
| | 'A, MAXI |
| | 'A pipe, Density of water in the downcomer/lower plenum, gc Conversi'on constant. |
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| | ii I |
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| | Modification Equation (I) was modified as 2 2 I/2 (1A A MAX c A B F spill (2) spill The difference between Equations (I) and (2) is the first term of the numerator on the right hand side of the equations. This change uses the real head term instead of the fixed head term while the mixture level is in the span of the cold .leg. |
| | C. Reasons for the Modification As indicated in the previous section, this change reflects a better physical representation than the model described in Reference l. |
| | However, a more important reason for this modification is to remove low-amplitude flow oscillations introduced by the discontinuity of the fixed head in the old model. |
| | D. Impact of the Spil.lage Model Change on PCT The change in downcomer spillage head term has the possibility to affect PCT through two effects -- reflood rate and two-phase level. |
| | Comparison of the reflood rates for cases without and with the change in head term shows that the small oscillations in the reflood rate are removed. However, the reflood rate selected for subsequent use is not changed; therefore, there is no change in PCT from this effect. The change in the head term for downcomer spillage also eliminates oscillations in the two phase-level but does not change the base two-phase level. Elimination of the oscillations reduces the uncertainty in the two-phase level selected for the next step in the analysis. However, due: to the small'ensitivity of the C-E methodology to two-phase level changes,.the change in the two-phase |
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| | il ll, ! |
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| | level due to the change in the head term for spillage has no effect on PCT. |
| | 4.0 Conclusions The change to COMPERC-II has the potential to affect the PCT by changing the reflood rate or the two-phase level. However, an evaluation, of the reflood rates and effect of the two-phase level for cases before and after the change in head term for the downcomer spillage shows that there is no change in PCT. |
| | The cummulative change in. PCT for large break LOCA including that from the previous annual report, Reference 5, is a reduction of, less than 1 F. |
| | There have been no changes in'he small break LOCA results to date. |
| | Therefore, there was no significant change in the sense of CFR 50.46 in 1989 and no action beyond the submission of this report is needed. |
| | 5.0 References |
| | : 1. "Emergency Core Cooling System; Revisions to Acceptance Criteria," |
| | 10CFR50, Federal Register,, Vol. 53, No. 180, September 16, 1988. |
| | : 2. CENPD-134P, "COMPERC-II, A Program for Emergency-Refil.l-Reflood of the Core," August, 1974. |
| | : 3. CENPD-134P, Supplement 1,. "COMPERC-II, A Program for Emergency |
| | 'Refill-Reflood of the Core (Modifications)," February, 1975. |
| | : 4. CENPD-134, Supplement 2, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June, 1985. |
| | : 5. CENPD-279, Annual Report on C-E .ECCS Codes and Methods for 10CFR50.46, April, 1989., |
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Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML17311B3441995-09-30030 September 1995 Safety Analysis Rept for Use of Advanced Zr Based Cladding Matl in PVNGS Unit 2 Batch J Demonstration Fuel Assemblies. ML17310B2471994-02-28028 February 1994 Nonproprietary Analysis of 137 Degree Capsule from APS Palo Verde Unit 2 Reactor Vessel Radiation Surveillance Program. ML20128E0741992-07-31031 July 1992 Nonproprietary Pvngs,Unit 1 EOC 3 Exam Rept ML17305B5141991-04-30030 April 1991 Suppl 2 to, Annual Rept on C-E ECCS Codes & Methods for 10CFR50.46. ML17305A6051990-02-28028 February 1990 Annual Rept on C-E ECCS Codes & Methods for 10CFR50.46, Suppl 1 ML19327C1951989-10-31031 October 1989 Nonproprietary Palo Verde Nuclear Generating Station-Unit 1 End-of-Cycle 2 Fuel Exam Rept. ML17223A6971989-04-30030 April 1989 Annual Rept on C-E ECCS Codes & Methods for 10CFR50.46. ML20155G7101988-05-31031 May 1988 End-Of-Cycle 1 Surveillance Fuel Exam Rept ML20150E6661988-02-29029 February 1988 Charging Pump Gas Binding & Cracked Block Evaluation Rept ML20235M2211987-07-13013 July 1987 Nonproprietary Rev 01-NP to Modified Statistical Combination of Uncertainties ML20236B6211987-06-30030 June 1987 Structural & Seismic Evaluation of Palo Verde Charging Pump Block ML20155J4911986-05-31031 May 1986 Rev 0 to Core Protection Calculator/Control Element Assembly Calculator Software Mods for CPC Improvement Program Reload Data Block ML20155J5051986-05-31031 May 1986 Rev 1 to Functional Design Requirements for Control Element Assembly Calculator ML20155J5231986-05-31031 May 1986 Rev 1 to Functional Design Requirement for Core Protection Calculator ML20151Z0521986-01-31031 January 1986 Rev 3 to CEN-39(A)-NP, CPC Protection Algorithm Software Change Procedure ML20151Z0641986-01-31031 January 1986 Rev 0 to CEN-323-NP Reload Data Block Constant Installation Guidelines ML20210E4871985-12-31031 December 1985 Comprehensive Vibration Assessment Program for Palo Verde Generating Station Unit 2 (Sys 80 Nonprototype-Category 1), Evaluation of Pre-Core Hot Functional Insp Program, Preliminary Rept ML20134N3021985-07-31031 July 1985 Rev 00 to Functional Design Requirement for Core Protection Calculator ML20134N2941985-07-31031 July 1985 Rev 00 to Functional Design Requirement for Control Element Assembly Calculator ML20111C1331985-02-28028 February 1985 Nonproprietary Rev 1-NP to, Response to NRC Questions on Core Protection Calculator Software,Palo Verde Nuclear Generating Station,Unit 1 ML20111C1171985-01-31031 January 1985 Rev 1 to Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (Sys 80 Prototype),Evaluation of Predictions & Pre-Core Hot Functional Measurement & Insp Programs ML17298B5401984-11-30030 November 1984 Summary Rept on Operability of Shutdown Cooling Sys Relief Valves for Palo Verde Units 1,2 & 3. ML17298B5391984-11-30030 November 1984 Summary Rept on Design Basis of Shutdown Cooling Sys Relief Valves for Cessar Sys 80. ML20100J2261984-09-30030 September 1984 Rev 2 to Core Protection Calculation/Control Element Assembly Calculator Sys Phase I Software Verification Test Rept ML20100J2541984-09-30030 September 1984 Rev 2 to Core Protection Calculation/Control Element Assembly Calculator Sys Phase II Software Verification Test Rept ML20100J2821984-09-30030 September 1984 Rev 1 to Palo Verde Nuclear Generating Station Cycle 1 Core Protection Sys & Control Element Assembly Calculator Data Base Listing ML17298B4991984-08-31031 August 1984 Errata to CEN-267(V)-NP & Rev 1-NP, Final Rept on Performance Evaluation of Palo Verde Control Element Assembly Shroud. ML20237J3131984-08-31031 August 1984 Rev 1 to Final Rept on Palo Verde Nuclear Generating Station Reactor Coolant Pumps ML20238E7341984-08-31031 August 1984 Nonproprietary Rept on Palo Verde Nuclear Generating Station LPSI Pumps Failure to Start. Related Documentation Encl ML20094F6321984-07-31031 July 1984 Rev 0 to Calculation of Trip Setpoint Values,Plant Protection Sys ML20080A7571984-01-31031 January 1984 Nonproprietary Interim Rept on Performance Evaluation of Palo Verde Control Element Assembly Shroud ML20100M1261984-01-31031 January 1984 Interim Rept on Palo Verde Nuclear Generating Station Reactor Coolant Pumps ML20079Q9031984-01-24024 January 1984 Nonproprietary Rept on Palo Verde Unit 1 Safety Injection Nozzle Thermal Liner ML20081A1211983-09-30030 September 1983 Rev 1 to Suppl 3 of Probabilistic Risk Assessment of Effects of PORVs on Depressurization & Dhr ML20071E7151983-01-31031 January 1983 Core Protection Calculation & Control Element Assembly Calculator Data Base Document. Proprietary Info Deleted ML20054L7051982-07-0202 July 1982 Nonproprietary Comprehensive Vibration Assessment Program for Prototype Sys 80 Reactor Internals,Palo Verde Nuclear Generating Station Unit 1 ML17297B1681981-12-31031 December 1981 Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for Palo Verde 1,2 & 3 Reactor Vessels. 1995-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17300B3811999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pvngs,Units 1,2 & 3.With 991007 Ltr ML17300B3271999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Pvngs,Units 1,2 & 3 ML17313B0751999-08-27027 August 1999 LER 99-002-00:on 990730,test Mode Trip Bypass for EDG Output Breakers Not Surveilled.Cause Under Investigation.Operations Personnel Conservatively Invoked SR 3.0.3 for SR 3.8.1.13. with 990827 Ltr ML17313B0611999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pvngs,Units 1,2 & 3.With 990810 Ltr ML17313B0191999-07-16016 July 1999 LER 99-005-00:on 990618,RT on Low DNBR Was Noted.Caused by Hardware Induced Calculation Error.Cr Operator Was Taken to Place Reactor in Stable Condition IAW Appropriate Operating Procedure ML17300B3151999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pvngs,Units 1,2 & 3.With 990714 Ltr ML17313A9921999-06-21021 June 1999 Special Rept:On 990525,RMS mini-computer Was Removed from Service to Implement Yr 2000 Mod & Was OOS Longer than 72 H Allowed.Caused by Planned Y2K Mods.Preplanned Alternate Sampling Program Was Initiated ML17313A9911999-06-18018 June 1999 Special Rept:On 990510,loose-part Detection Sys Channel 2 Was Declared Inoperable.Caused by Malfunction of Mineral Cable Connector to Accelerometer.Licensee Will Implement Modifications Which Will Enhance loose-part Detection Sys ML17313A9731999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pvngs,Units 1,2 & 3.With 990608 Ltr ML17313A9281999-05-0707 May 1999 LER 99-004-00:on 990408,PSV Lift Pressures Were Outside of TS Limits.Caused by Lift Pressure Setpoint Drift.Psvs Have Been Tested,Disassembled,Inspected,Reassembled & Certified at Wyle Labs ML17313A9201999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pvngs,Units 1,2 & 3.With 990512 Ltr ML17313A8951999-04-14014 April 1999 LER 99-003-00:on 990317,required Surveillance Requirement Not Completed Due to Deficient Procedure,Was Determined. Caused by Cognitive Personnel Error.St Procedures Revised to Require Chiller to Be Operating & Oil Temperature Checked ML17313A8921999-04-13013 April 1999 LER 98-003-01:on 980902,discovered That MSSV as-found Lift Pressures Were Outside TS Limits.Caused by Bonding of Valve Disc to Nozzle Seat.Affected Valves Were Adjusted,Retested & Returned to Svc ML17313A8891999-04-0909 April 1999 LER 99-001-00:on 990310,RT on High Pressurizer Pressure Was Noted.Caused by Loss of Heat Removal.Cr Supervisor Was Removed from Shift Duties for Diagnostics Skills Training. with 990409 Ltr ML17300B3071999-03-31031 March 1999 Seismic Portion of Submittal-Only Screening Review of Palo Verde Nuclear Generating Station Units Ipeee. ML17313A8801999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pvngs,Units 1,2 & 3.With 990412 Ltr ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207H7471999-03-10010 March 1999 1999 Emergency Preparedness Exercise 99-E-AEV-03003 ML17313A8361999-03-0101 March 1999 LER 99-001-00:on 990103,TS Violation for Power Dependent Insertion Limit Alarm Being Inoperable.Caused by Personnel Error.Revised Procedure to Clarify How Computer Point Is to Be Returned to Scan Mode.With 990302 Ltr ML17313A8501999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Palo Verde Nuclear Generating Station.With 990311 Ltr ML17313A7791999-02-0505 February 1999 Safety Evaluation Accepting Licensee Rev to Emergency Plan That Would Result in Two Less Radiation Protection Positions Immediatelu Available During Emergencies ML17313A8061999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Pvngs,Units 1,2 & 3.With 990218 Ltr ML17313A7701999-01-15015 January 1999 LER 96-008-00:on 960507,inadequate Procedure Results in Nuclear Power Channels Not Calibrated During Power Ascension Tests Occurred.Caused by Deficient Procedure.Procedure Revised ML17313A7381998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Palo Verde Nuclear Generating Station,Units 1,2 & 3.With 990113 Ltr ML20206H2101998-12-31031 December 1998 SCE 1998 Annual Rept ML17313A7031998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Pvngs,Unit 1,2 & 3. with 981209 Ltr ML17313A6701998-11-0404 November 1998 Rev 2 to PVNGS Unit 2 Colr. ML17313A6741998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Pvngs,Units 1,2 & 3.With 981109 Ltr ML17313A6611998-10-24024 October 1998 LER 98-008-00:on 980729,EQ of Electrical Connectors Were Not Adequately Demonstrated.Caused Because Test Was Conducted with Only Single Lv Connector & Without Fully Ranged Inputs. Revised EQ Requirements ML17313A6561998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for PVNGS Units 1,2 & 3.With 981007 Ltr ML17313A5961998-09-14014 September 1998 LER 98-002-00:on 980814,B Train H Recombiner Was Noted Inoperable Due to cross-wired Power Receptacle.Cause of Event Is Under investigation.Cross-wired Power Supply Receptacle for B Train H Recombiner Was re-wired ML17313A5761998-09-0808 September 1998 LER 98-003-01:on 980113,discovered That One Channel of RWT Level Sys Had Failed High.Caused by Water Intrusion Into Electrical Termination Pull Box.Weep Holes Were Drilled Into Bottoms of Pull Boxes Nearest Level Transmitters ML17313A5591998-08-28028 August 1998 LER 98-001-00:on 980730,entered TS 3.0.3 Due to Safety Injection Flow Instruments Being Removed from Svc.Caused by Personnel Error.Transmitters Were Unisolated & Returned to svc.W/980828 Ltr ML20151S0941998-08-21021 August 1998 Rev 6 to COLR for PVNGS Unit 3 ML20151S0861998-08-21021 August 1998 Rev 4 to COLR for PVNGS Unit 1 ML20151S0901998-08-21021 August 1998 Rev 1 to COLR for PVNGS Unit 2 ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML17313A5401998-08-13013 August 1998 Special Rept:On 980715,declared PASS Inoperable.Caused by Failure of Offgas Flush/Purge Control Handswitch HS0101. Handswitch Replaced & Post Maintenance Retesting Was Initiated ML17313A5301998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Pvgns,Units 1,2 & 3.W/980812 Ltr ML17313A5201998-07-30030 July 1998 LER 98-004-00:on 980630,personnel Discovered That Pressure Safety Valve Had Not Received Periodic Set Pressure Test for ASME Class 1 Pressure Safety Valve.Caused by Personnel Error.Pressure Safety Valve reviewed.W/980730 Ltr ML17313A5791998-07-0707 July 1998 to PVNGS SG Tube ISI Results for Seventh Refueling Outage Mar & Apr 1998. ML17313A5001998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Palo Verde Nuclear Generating Station,Units 1,2 & 3.W/980710 Ltr ML17313A4671998-06-19019 June 1998 LER 98-007-00:on 980520,CR Personnel Observed Flow & Pressure Perturbations on Chemical & Vol Control Sys Letdown Sys.Caused by Cyclic Fatigue Due to Dynamic Pressure Transients.Unit Letdown Piping Replaced ML17313A4521998-06-19019 June 1998 Rev 5 to COLR for Pvngs,Unit 3. ML17313A4501998-06-19019 June 1998 Rev 4 to COLR for Pvngs,Unit 3. ML17313A4131998-06-0505 June 1998 LER 98-006-00:on 980507,determined That Plant Was Outside Design Basis Due to SI Discharge Check Valve Reverse Flow. Check Valve Was Disassembled,Examined & Reassembled, Whereupon Valve Met Acceptance Criteria ML17313A4211998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Pvngs,Units 1,2 & 3.W/980609 Ltr ML17313A3951998-05-26026 May 1998 LER 98-005-00:on 980428,noted That Required Response Time Testing Had Not Been Performed.Caused by Personnel Error. Coached I&C Personnel Responsible for Reviewing Work Authorization Documentation ML17313A3691998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for PVNGS.W/980412 Ltr ML17313A3251998-04-0101 April 1998 LER 98-004-00:on 980304,safety Valves as-found Pressures Out of Tolerance.Cause of Event Is Under Investigation.Three Mssv'S & Psv Will Be Replaced W/Refurbished & Recertified Valves During Refueling Outage U1R7 1999-09-30
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Text
CENPD - 279 SUPPLEMENT 1 ANNUAL REPORT ON
'C-E ECCS CODES AND METHODS FOR 10CFR50.46 FEBRUARY. 1990 9003260577 900228 PDR AQOCK 05000~28 P
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LEGAL NOTICE This report was prepared as an account of, work sponsored .by Combustion Engineering, Inc. 'Neither. Combustion Engineering nor any person acting on its .behalf:
A. makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose. or merchantabil,ity, wi.th respect to the accuracy, completeness, or usefulness of the information contained in this report, or 'that the use- of any information, apparatus, method, or process disclosedin
.this report may not infringe privately owned rights; or B. assumes any l,i'abilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed 'in this report.
CENPD - 279 SUPPLEMENT, 1 ANNUAL REPORT ON C-E ECCS CODES AND METHODS FOR 10CFR50.46 TRANslENT METHoos ANo LOCA NUCLEAR FUEL ENGlNEERING FEBRUARY 1990
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Abstract This report describes changes and errors in the Combustion Engineering codes and anal'ysis methodology for ECCS analysis in 1989 per the requirements of 10CFR50.46. For this reporting period only one computer code had reportable changes or errors. The corrections and changes did not affect the peak cladding temperature. The,cummulative temperature change for large break LOCA is a reduction of less than. 1 F. No changes or errors..that affect the peak cladding temperature for small break LOCA have occured. Per the:.cri.teria of 10CFR50.46, no action .beyond this annual report is required.
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Table of Contents 1.0 'Introduction 2.0 Codes for ECCS Evaluation 3.0 Error Corrections. and Model Changes in Computer Codes
- 3. 1 COMPERC-II 4',0 Conclusions 5.0 References
0 0 1.0 Introduction This report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis. The revision to the ECCS Acceptance Criteria spells out reporting requirements and actions required when errors are corrected or changes are made in an evaluation model or in the application of a model for an operating licensee or construction permittee of a nuclear power plant.
The action requirements in g 50.46(a)(3) are:
- l. Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For, this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature (PCT) different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 F.
- 2. For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in g 50.4.. This report is,to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a revised evaluation correcting minor errors is approved by the NRC staff.
- 3. If the change or error is significant, the applicant or licensee shall provide this report. within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with g 50.46
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requirements. This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC.
For those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the proposed schedule.
- 4. Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in paragraph (b) of g 50.46 is a reportable event as described in gg 50.55(e), 50.72 and 50.73. The affected applicant or licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with g 50.46 requirements.
This report documents all the changes made to the presently licensed C-E LOCA analysis models and methodology which have not been reviewed by the NRC staff. This is specifically to satisfy the requirements described in the second item above.
2.0 Codes for ECCS Evaluation C-E uses several digital computer codes for ECCS analysis that are described in topical reports, are licensed by the NRC and are covered by the provisions of CFR 50.46. Those for large break LOCA .calculations are: CEFLASH-4A, COMPERC-II, PARCH, STRIKIN-II, and COMZIRC. CEFLASH-4AS is used in conjunction with COMPERC-II, STRIKIN-II, and PARCH for small break LOCA calculations.
3.0 Error Corrections and Model Changes in Computer Codes
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This section discusses all error corrrections or model changes to the licensed codes which may affect the calculated PCT. Only the COMPERC-II for a large break has been changed in 1989. No changes to analysis procedures have been made since the, last approved submittal to the NRC.
~l 3 ..I., COMPERC- II A. Code Description COHPERC-II is a FORTRAN digital computer program which is used by Combustion Engineering, Inc. to calculate the core refill and reflood transient portion of a PWR loss of coolant accident (LOCA).
A detailed code. description is presented in References 2 through 4.
B. Model Change in COMPERC-II for SI. Spillage The model for the spil.lage calculation .in COMPERC-II'has been changed .to reflect a more realistic physical representation. The change .is as described below:
Nd1 (Pg 10 f'Rf 2)'f ZA, HAX ZA ZA, MAXI c'
2 A
2 I/2 A'MAXI A MAX B F spill spill where Wspill Rate of water spillage out of the break, spill Loss coefficient for,the spillage of water out of the break,.
B F Flow, area in the core, ZA Height of the water in the downcomer, A,MAX Distance between bottom of core and .bottom of inlet pipe, Dis'tance between bottom of core and top of inlet
'A, MAXI
'A pipe, Density of water in the downcomer/lower plenum, gc Conversi'on constant.
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Modification Equation (I) was modified as 2 2 I/2 (1A A MAX c A B F spill (2) spill The difference between Equations (I) and (2) is the first term of the numerator on the right hand side of the equations. This change uses the real head term instead of the fixed head term while the mixture level is in the span of the cold .leg.
C. Reasons for the Modification As indicated in the previous section, this change reflects a better physical representation than the model described in Reference l.
However, a more important reason for this modification is to remove low-amplitude flow oscillations introduced by the discontinuity of the fixed head in the old model.
D. Impact of the Spil.lage Model Change on PCT The change in downcomer spillage head term has the possibility to affect PCT through two effects -- reflood rate and two-phase level.
Comparison of the reflood rates for cases without and with the change in head term shows that the small oscillations in the reflood rate are removed. However, the reflood rate selected for subsequent use is not changed; therefore, there is no change in PCT from this effect. The change in the head term for downcomer spillage also eliminates oscillations in the two phase-level but does not change the base two-phase level. Elimination of the oscillations reduces the uncertainty in the two-phase level selected for the next step in the analysis. However, due: to the small'ensitivity of the C-E methodology to two-phase level changes,.the change in the two-phase
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level due to the change in the head term for spillage has no effect on PCT.
4.0 Conclusions The change to COMPERC-II has the potential to affect the PCT by changing the reflood rate or the two-phase level. However, an evaluation, of the reflood rates and effect of the two-phase level for cases before and after the change in head term for the downcomer spillage shows that there is no change in PCT.
The cummulative change in. PCT for large break LOCA including that from the previous annual report, Reference 5, is a reduction of, less than 1 F.
There have been no changes in'he small break LOCA results to date.
Therefore, there was no significant change in the sense of CFR 50.46 in 1989 and no action beyond the submission of this report is needed.
5.0 References
- 1. "Emergency Core Cooling System; Revisions to Acceptance Criteria,"
10CFR50, Federal Register,, Vol. 53, No. 180, September 16, 1988.
- 2. CENPD-134P, "COMPERC-II, A Program for Emergency-Refil.l-Reflood of the Core," August, 1974.
- 3. CENPD-134P, Supplement 1,. "COMPERC-II, A Program for Emergency
'Refill-Reflood of the Core (Modifications)," February, 1975.
- 4. CENPD-134, Supplement 2, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June, 1985.
- 5. CENPD-279, Annual Report on C-E .ECCS Codes and Methods for 10CFR50.46, April, 1989.,
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