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Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20248D6011998-03-31031 March 1998 Suppl 9 to CENPD-279, Annual Rept on Abb CE ECCS Performance Evaluation Models ML20203H6701998-01-31031 January 1998 Rev 0 to non-proprietary Version of BAW-10232, OTSG Repair Roll Qualification Rept (Including Hydraulic Expansion Evaluation) ML20199G9531998-01-31031 January 1998 Non-proprietary Alternate Repair Criteria for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once- Through Sgs ML20211N4921997-09-30030 September 1997 Rev 1 to SIR-94-080-A, Relaxation of Reactor Coolant Pump Flywheel Insp Requirements ML20135F3681996-11-30030 November 1996 Non-proprietary Final Rept Repair of 3/4 O.D. SG Tubes Using Leak Tight Sleeves ML20112E8491996-02-28028 February 1996 Suppl 7 to Annual Rept of Abb C-E ECCS Performance Evaluation Models ML17353A4341995-10-31031 October 1995 Rev 1 to BAW-2245, Initial Rt of Linde 80 Welds Based on Fracture Toughness in Transition Range. ML17264A1181995-07-31031 July 1995 Response to Part (1) of GL 92-01,Rev 1,Suppl 1. ML20083Q9261995-05-30030 May 1995 Joint Applications Rept for Safety Injection Tank Aot/Sti Extension ML20083Q9861995-05-30030 May 1995 Joint Applications Rept for Emergency Diesel Generators AOT Extension ML20085H3221995-02-28028 February 1995 Suppl 6 to Topical Rept CENPD-279, Annual Rept on Abb CE ECCS Performance Evaluation Models,Final Rept,Ceog Task 865, Dtd Feb 1995 ML20073D4561994-09-30030 September 1994 Verification of Cecor Coefficient Methodology for Application to PWRs of Entergy Sys ML20069D4261994-02-28028 February 1994 Annual Rept on C-E ECCS Codes & Methods for 10CFR50.46 ML20063C7571993-12-31031 December 1993 Qualification of Reactor Methods for Pressurized Water Reactors of Entergy Sys ML20081K9821991-05-31031 May 1991 Final Rept on Reactor Vessel App G Pressure-Temp Limits for Arkansas Nuclear One Unit 2 for 21 Efpys ML20067A5241990-12-31031 December 1990 Final Submittal in Response to NRC Bulletin 88-011, 'Pressurizer Surge Line Thermal Stratification.' ML20042F3391990-04-30030 April 1990 Suppl 1 to Responses to Questions on C-E Rept CEN-386-P, 'Verification of Acceptability of 1-Pin Burnup Limit of 60 Mwd/Kg for C-E 16X16 PWR Fuel.' ML20059M4481990-03-31031 March 1990 Unit 1 Pressure-Temp Limits for 15 Efpy ML20247E3231989-06-30030 June 1989 Nonproprietary Verification of Acceptability of 1-Pin Burnup Limit of 60 Mwd/Kg for C-E 16x16 PWR Fuel ML20245G7121989-05-31031 May 1989 Submittal in Response to NRC Bulletin 88-11, 'Pressurizer Surge Line Thermal Stratification' ML20246N3841989-04-30030 April 1989 Analysis of Capsule ANI-C,Arkansas Power & Light Co, Arkansas Nuclear One,Unit 1,Reactor Vessel Matl Surveillance Program ML20211A1031986-09-30030 September 1986 Small Break LOCA Analysis for B&W 177FA Lowered Loop Plants in Response to NUREG-0737,Item II.K.3.31 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 ML20202B1131986-04-30030 April 1986 Nonproprietary Suppl 1,Rev 3, CPC Protection Algorithm Software Change Procedure 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 ML20135A5571985-08-31031 August 1985 B&W Owners Group Cavity Dosimetry Program 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 ML20129F9301985-07-31031 July 1985 Cycle 5 Shoulder Gap Evaluation ML20100J1551985-03-31031 March 1985 Nonproprietary Typical Data Base Constants for Arkansas Nuclear One Unit 2 ML20100J1801985-03-31031 March 1985 Nonproprietary Rev 0 to Arkansas Nuclear One - Unit 2 Core Protection Calculator Sys Phase I Design Qualification Test Rept ML20100J1991985-03-31031 March 1985 Nonproprietary Rev 0 to Arkansas Nuclear One - Unit 2 Core Protection Calculator & Control Element Assembly Calculator Data Base Listing ML20100J2221985-03-31031 March 1985 Nonproprietary Rev 1 to Core Protection Calculator/Control Element Assembly Calculator Sys Phase II Software Verification Test Rept ML20101U3551984-12-31031 December 1984 Nonproprietary Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2 ML20106E3251984-10-31031 October 1984 Nonproprietary Once-Through Steam Generator Mechanical Sleeve Qualification ML20107N0221984-10-31031 October 1984 Nonproprietary CPC Methodology Changes for Arkansas Nuclear One Unit 2 Cycle 5 ML20094J7101984-07-31031 July 1984 Analyses of Capsule AN1-A,Arkansas Power & Light Co, Arkansas Nuclear One,Unit 1,Reactor Vessel Matl Surveillance Program ML20098F1361984-04-30030 April 1984 Thermal-Hydraulic Crossflow Applications ML20083G5321983-11-30030 November 1983 Cycle 4 Shoulder Gap Evaluation ML20087N2541983-11-30030 November 1983 Nonproprietary Shoulder Gap Data Taken on Batch D Assemblies After Cycle 3. Info Deleted ML20066D2241982-03-31031 March 1982 Effects of Vessel Head Voiding During Transients & Accidents in C-E Nsss. Portions Intentionally Deleted Due to Lack of Relevancy to NUREG-0737,Item II.K.2.17.Util Did Not Participate in Development of Deleted Sections ML20039F8681981-12-31031 December 1981 Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for AR Nuclear One- Unit 2 Reactor Vessel. ML20009F2021981-07-31031 July 1981 Nonproprietary Version of Response to Questions on Documents Supporting ANO-2,Cycle 2,License Submittal, Amend 2-NP 1999-07-29
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c 0CAN109902, Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20216J6271999-09-27027 September 1999 Rev 0 to CALC-98-R-1020-04, ANO-1 Cycle 16 Colr ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 0CAN099907, Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety 0CAN089904, Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates 0CAN079903, Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20196A0191999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20195D1991999-05-28028 May 1999 Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 0CAN059903, Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205R6351999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ano,Units 1 & 2. with ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) 0CAN039904, Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with ML20212G6381999-02-25025 February 1999 Ano,Unit 2 10CFR50.59 Rept for 980411-990225 ML20203E4891999-02-11011 February 1999 Rev 1 to 97-R-2018-03, ANO-2,COLR for Cycle 14 ML20199F0351998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Ano,Units 1 & 2 ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 0CAN129805, LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With1998-12-11011 December 1998 LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20198D2441998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Ano,Units 1 & 2. with ML20199F7401998-11-16016 November 1998 Rev 9 to ANO-1 Simulator Operability Test,Year 9 (First Cycle) ML20195B4801998-11-0707 November 1998 Rev 20 to ANO QA Manual Operations ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program 0CAN119808, Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with ML20197H0741998-10-29029 October 1998 Rev 1 to Third Interval ISI Program for ANO-1 ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML17335A7641998-10-22022 October 1998 LER 98-004-00:on 980923,inadvertent Actuation of Efs Occurred During Surveillance Testing.Caused by Personnel Error.Personnel Involved with Event Were Counseled & Procedure Changes Were Implemented.With 981022 Ltr ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 0CAN109806, Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with ML20154E2171998-09-28028 September 1998 Follow-up Part 21 Rept Re Defect with 1200AC & 1200BC Recorders Built Under Westronics 10CFR50 App B Program. Westronics Has Notified Bvps,Ano & RBS & Is Currently Making Arrangements to Implement Design Mods 0CAN099803, Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with ML20237B7671998-08-19019 August 1998 ANO REX-98 Exercise for 980819 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 ML20236X2351998-08-0505 August 1998 Part 21 Rept Re Defect Associated W/Westronics 1200AC & 1200BC Recorders Built Under Westronics 10CFR50,App B Program.Beaver Valley,Arkansas Nuclear One & River Bend Station Notified.Design Mod Is Being Developed 0CAN089804, Monthly Operating Repts for July 1998 for Ano,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Ano,Units 1 & 2 ML20196C7831998-07-30030 July 1998 Summary Rept of Results for ASME Class 1 & 2 Pressure Retaining Components & Support for ANO-1 ML20155H7161998-07-15015 July 1998 Rev 1 to 96-R-2030-02, Revised Reactor Vessel Fluence Determination ML20236R0531998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Ano,Units 1 & 2 ML20249B7791998-06-22022 June 1998 Part 21 Rept Re Findings,Resolutions & Conclusions Re Failure of Safety Related Siemens 4KV,350 MVA,1200 a Circuit Breakers to Latch Closed ML20249B5091998-06-15015 June 1998 SG ISI Results for Fourteenth Refueling Outage 1999-09-30
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CENPD-279 SUPPLEMENT 7 i
- ANNUAL REPORT ON ABB CE ECCS ,
- PERFORMANCE EVALUATION MODELS FINAL REPORT 1
1 February 1996 0 Copyright 1996 Combustion Engineering, Inc. All rights reserved
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ABB Combustion Engineering Nuclear Operations D0 ADO 05 0 13 8 I oI 7% 9 D ASEA BROWN BOVERI
LEGAL NOTICE This report was prepared as an account of work sponsored by ABB Combustion Engineering.
Neither Combustion Engineering, Inc. nor any person acting on its behalf: '
A. makes any warranty or represen% tion, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any infonnation, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.
Combustion Engineering, Inc.
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ARSTRACT l This report describes changes and errors in the ABB Combustion Engineering evaluation models I for ECCS analysis in 1995 per the requirements of 10CFR50.46. For this reporting period, one ;
error in the input processing for the COMPERC-II refill /reflood code for large break LOCA ,
analysis was found and corrected. No other changes were made to the ABB CE evaluation models for the large break, small break or post-LOCA long term cooling calculations. l l
Correction of the error in COMPERC-II had no effect on the cladding temperature (PCT) for large break LOCA. The sum of the absolute magnitudes of the PCT changes for large break LOCA from all reports to date continues to be less than 1 'F. No change occured in the PCT for small break LOCA or post-LOCA long term cooling. Per the criteria of 10CFR50.46, no action beyond this annual report is required.
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TABLE OF CONTENTS l l
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1.0 INTRODUCTION
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l 2.0 ABB CE CODES USED FOR ECCS EVALUATION 3 3.0 EVALUATION MODEL CHANGES AND ERROR CORRECTIONS 4 t 3.1 COMPERC-II for Large Break LOCA 4 3.1.1 Code Description i 3.1.2 Error in COMPERC-II i
( 3.1.3 Correction of COMPERC-II Code Error l
! 3.1.4 Impact of COMPERC-II Error on PCT
4.0 CONCLUSION
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5.0 REFERENCES
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1.0 INTRODUCTION
This report addresses the NRC requirement to report changes or errors in ECCS performance evaluation models. The ECCS Acceptance Criteria, Reference 1, 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 plav ,
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l The action requirements in 10CFR50.46(a)(3) are- 1 l
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- 1. 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 j 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 j 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
'a a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 F.
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- 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 O
10CFR50.4.
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- 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 10CFR50.46 requirements. This schedule may be developed using an l 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 10CFR50.46 is a reportable event as described in 10CFR50.55(e),50.72 and 50.73. The Miected applicant or licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with 10CFR50.46 requirements.
This report documents all the errors corrected in and/or changes to the presently licensed ABB CE ECCS performance evaluation models, made in the year covered by this report, which have not been reviewed by the NRC staff. This document is provided to satisfy the reporting requirements of the second item above. ABB CE reports for earlier years are given in References 2-8.
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F 2.0 ABB CE CODES USED FOR ECCS EVALUATION l ABB CE uses several digital computer codes for ECCS performance analysis that are described in topical reports, are licensed by the NRC, and are covered by the provisions of 10CFR50.46.
Those for large break LOCA calculations are CEFLASH-4A, COMPERC-II, HCROSS, PARCH, STRIKIN-II, and COMZIRC. CEFLASH-4AS is used in conjunction with COMPERC-II, SBUKIN-II, and PARCH for small break LOCA calculations. The codes for post-LOCA long ,
term cooling analysis are BORON, CEPAC, NATFLOW, and CELDA. l l
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3.0 EVALUATION MODEL CHANGES AND ERROR CORRECTIONS This section discusses all error corrections and model changes to the ABB CE ECCS performance evaluation models which may affect the calculated PCT. In 1995 an error in the input processing for one computer code used in the large break LOCA evaluation model was corrected. The nature of this error and the steps taken to resolve it are described below.
3.1 COMPERC-II for Larce Break LOCA 3.1.1 Code Description COMPERC-II calculates the reactor cooling system (RCS) hydraulic response during the refill /reflood portion of a large break LOCA transient. Models are provided in the code for the hydraulic behavior c f the NSSS, addition and removal of fluid, core heat transfer, containment pressure, related systems, and properties. It also calculates the reflood heat transfer coefficient for the claddi s using a FLECHT-based correlation.
3.1.2 Error in COMPERC-II The error identified is in the input processing for the containment pressure module of the code.
The code documentation, Reference 9, describes the maximum number of entries permitted for each input array that is used in the containment pressure module. However, the code does not
, check to ensure that the number of entries specified for each input array does not exceed the maximum number allowed.
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t' An error can occur if the number of entries specified exceeeds the maximum number supported.
For the FORTRAN language, use of more entries in an array than the number of entries defined
. .by the DIMENSION statement for the array overwrites the information in subsequent memory locations. This has the possibility of causing erroneous results depending on the specific array involved.
3.1.3 Correction of COMPERC-II Code Error ,
The error was addressed by revising the administrative procedures controlling use of the COMPERC-II code.
3.1.4 Impact of COMPERC-II Error on PCT No ECCS performance analyses using the ABB CE large break LOCA evaluation model are impacted by this error. Consequently, there is no impact on the calculated PCT.
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4.0 CONCLUSION
S One error was found and corrected in the COMPERC-II computer code used for large break LOCA analysis during 1995. There was no change in the PCT as a result of correcting this error.
No other changes to the models and methods or corrections of errors were made in 1995. The sum of the absolute magnitudes of the changes in PCT calculated using the C-E ECCS evaluation models, including those from previous annual reports, References 2-8, remains less than 1 *F.
Based on the results reported here, there was no significant change in the sense of 10CFR50.46 in 1995 and no action beyond the submission of this report is needed.
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5.0 REFERENCES
- 1. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear
, Power Reactors," Code of Federal Regulations, Title 10, Part 50, Section 50.46.
- 2. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, April, 1989.
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! 3. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, Supplement 1, February,1990.
- 4. " Annual Repert on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, Supplement 2, April,1991.
l S. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, j Supplement 3, April,1992.
- 6. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, Supplement 4, April,1993.
- 7. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279,
! Supplement 5, February,1994.
l 8. " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," CENPD-279, Supplement 6, February,1995,
- 9. "COMPERC-II, A Program for Emergency-Refill-Reflood of the the Core," CENPD-134 P, August,1974 COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modifications),"
CENPD-134 P, Supplement 1, February,1985.
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