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Category:TEXT-SAFETY REPORT
MONTHYEARML20216G0111999-09-30030 September 1999 Year 2000 Readiness in U.S. Nuclear Power Plants ML20206N2191999-04-30030 April 1999 Operator Licensing Examination Standards for Power Reactors ML20205A5291999-03-31031 March 1999 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1998.(White Book) ML20211K2851999-03-31031 March 1999 Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance ML20205A5991999-03-31031 March 1999 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1998.(White Book) 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 ML20203D0541999-01-31031 January 1999 Fire Barrier Penetration Seals in Nuclear Power Plants ML20155A9281998-10-31031 October 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1998.(White Book) ML20154C2081998-09-30030 September 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1998.(White Book) ML15261A4681998-09-0404 September 1998 Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively ML20203A1521998-07-31031 July 1998 Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment ML20153D3371998-07-31031 July 1998 Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment ML20236S9771998-06-30030 June 1998 Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Pressurized Water Reactors ML20236S9681998-06-30030 June 1998 Evaluation of AP600 Containment THERMAL-HYDRAULIC Performance ML20236S9591998-06-30030 June 1998 Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Boiling Water Reactors ML20217Q7971998-05-0404 May 1998 Safety Evaluation Supporting Amends 227 & 201 to Licenses DPR-53 & DPR-69,respectively ML20247E3951998-04-30030 April 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1997.(White Book) ML20217F3801998-03-31031 March 1998 Risk Assessment of Severe ACCIDENT-INDUCED Steam Generator Tube Rupture ML20202J3051997-11-30030 November 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1997.(White Book) ML20197B0431997-11-30030 November 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1997.(White Book) ML20211L2931997-09-30030 September 1997 Aging Management of Nuclear Power Plant Containments for License Renewal ML20210K7801997-08-31031 August 1997 Topical Report Review Status ML20149G9431997-07-31031 July 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1997.(White Book) ML20210R2131997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the System 80+ Design.Docket No. 52-002.(Asea Brown Boveri-Combustion Engineering) ML20140F0801997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design.Supplement No. 1.Docket No. 52-001.(General Electric Nuclear Energy) ML20140J4301997-05-31031 May 1997 Safety Evaluation Report Related to the Department of Energy'S Proposal for the Irradiation of Lead Test Assemblies Containing TRITIUM-PRODUCING Burnable Absorber Rods in Commercial LIGHT-WATER Reactors ML20141J9391997-04-30030 April 1997 Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at North Carolina State University ML20141C2411997-04-30030 April 1997 Circumferential Cracking of Steam Generator Tubes ML20141A5791997-04-30030 April 1997 Proposed Regulatory Guidance Related to Implementation of 10 CFR 50.59 (Changes, Tests, or Experiments).Draft Report for Comment ML20137A2191997-03-31031 March 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1996.(White Book) ML20134L3601997-01-31031 January 1997 Standard Review Plan on Antitrust.Draft Report for Comment ML20134L3631997-01-31031 January 1997 Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance.Draft Report for Comment ML20138J2461997-01-31031 January 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1996.(White Book) ML20135D5711997-01-31031 January 1997 Operator Licensing Examination Standards for Power Reactors ML20133E9161996-12-31031 December 1996 License Renewal Demonstration Program: NRC Observations and Lessons Learned ML20149L8261996-10-31031 October 1996 Reactor Pressure Vessel Status Report ML20135A4981996-10-31031 October 1996 Historical Data Summary of the Systematic Assessment of Licensee Performance ML20128P4381996-10-0909 October 1996 Safety Evaluation Accepting Review of Cracked Weld Operability Calculations & Staff Response to NRC Task Interference Agreement ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML14183A6951995-09-18018 September 1995 Safety Evaluation Approving Relocation of Technical Support Ctr ML20236L5971994-12-29029 December 1994 SER in Response to 940314 TIA 94-012 Requesting NRR Staff to Determine Specific Mod to Keowee Emergency Power Supply Logic Must Be Reviewed by Staff Prior to Implementation of Mod ML20128Q0761994-11-0404 November 1994 Coordinating Group Evaluation,Conclusions & Recommendations ML20149H0671994-11-0404 November 1994 Safety Evaluation Supporting Amend 27 to Amended License R-103 ML20149G4281994-09-28028 September 1994 NRC Perspectives on Accident Mgt, Presented at 940928 Severe Accident Mgt Implementation Workshop in Alexandria, VA ML20149F7581994-08-25025 August 1994 Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program ML20149F4151994-08-0404 August 1994 Safety Evaluation Concluding That Unit 1 Can Be Safely Operated During Next Operating Cycle (Cycle 14) ML20149E8831994-08-0202 August 1994 Safety Evaluation Accepting Interim Relief Request IRR-03 Re Drywell Isolation Check Valves in Equipment Drain Lines & Reactor Equipment Closed Cooling Water Sys ML20248C5731994-07-19019 July 1994 SER Step 1 Review of Individual Plant Exam of External Fire Events for Millstone Unit 3 ML20059J4591994-01-25025 January 1994 Safety Evaluation Supporting Request for Relief from ASME Code Re Inservice Testing Requirements to Measure Vibration Amplitude Displacement ML20059H4991994-01-24024 January 1994 Safety Evaluation Accepting Revised Responses to IEB-80-04 Re MSLB Reanalysis 1999-09-30
[Table view]Some use of "" in your query was not closed by a matching "". Category:TOPICAL REPORT EVALUATION
MONTHYEARML20149F7581994-08-25025 August 1994 Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program ML20059L1061994-01-12012 January 1994 Draft Topical Rept Evaluation of B&Wog Rept 47-1223141-00, Integrated Plant Assessment Sys/Structure Screening.... Applicant for License Renewal That Refs B&Wog Sys Screening Methodology Will Be Required to Develop Own Procedures ML20059D1911993-12-30030 December 1993 Topical Rept Evaluation of RXE-91-005, Methodology for Reactor Core Response to Steamline Break Events ML20058P2181993-12-10010 December 1993 SER Accepting Siemens Nuclear Power Corp Submittal of Topical Rept EMF-92-081, Statistical Setpoint/Transient Methodology for W Type Reactors ML20058H9851993-11-26026 November 1993 Topical Rept Evaluation of WCAP-10216-P, Relaxation of Constant Axial Offset Control. Rept Acceptable ML20059H8481993-11-0202 November 1993 SER Accepting Proposed Changes in Rev 3 to OPPD-NA-8302-P, OPPD Nuclear Analysis,Reload Core Analysis Methodology, Neutronics Design Methods & Verification ML20134B4761993-10-30030 October 1993 Topical Rept Evaluation of Rev 3 to NP-2511-CCM Re VIPRE-01 Mod 2 for PWR & BWR Applications ML20058M9851993-09-30030 September 1993 SE of Topical Rept, Transient Analysis Methodology for Wolf Creek Generating Station ML20056G4171993-08-18018 August 1993 Topical Rept Evaluation of Rev 4 to OPPD-NA-8303, Transient & Accident Methods & Verification. Proposed Changes in Rev 4 Acceptable Except for Use of Cents Computer Code for Transient Analyses ML20056E9661993-08-0606 August 1993 Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containment Hydrogen Control ML20056E3811993-08-0505 August 1993 Safety Evaluation of RXE-89-002, Vipre-01 Core Thermal- Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications. Rept Is Acceptable for Ref in CPSES Core thermal-hydraulic Analyses ML20056E3961993-08-0505 August 1993 Safety Evaluation of RXE-90-006-P, Power Distribution Control Analysis & Overtemperature N-16 & Overpower N-16 Trip Setpoint Methodology. Methodology Acceptable ML20056E4681993-08-0505 August 1993 Supplemental Safety Evaluation for Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments. Change Requests Consistent & Compatible W/ 10CF50.44 & Acceptable ML20056E2571993-08-0505 August 1993 Corrected Safety Evaluation for Topical Rept RXE-91-001, Transient Analysis Methods for Commanche Peak Steam Electric Station Licensing Applications. Corrections Made to Second Sentense of Second Full Paragraph on Page Two ML20056D9921993-07-29029 July 1993 Topical Rept Evaluation of OPPD-NA-8301,Rev 5, Reload Core Analysis Methodology Overview. Proposed Changes in Rev 5 Acceptable ML20057A2661993-07-14014 July 1993 Topical Safety Evaluation of CEN-387-P, Pressurizer Surge Line Flow Stratification Evaluation. C-E Owners Group Analysis May Be Used as Basis for Licensees to Update plant- Specific Code Stress Rept for Compliance W/Bulletin 88-011 ML20056E1261993-06-29029 June 1993 Safety Evaluation of CENPD-382-P, Methodology for Core Designs Containing Erbium Burnable Absorbers. Rept Acceptable for Reload Licensing Applications of Both CE CE 14x14 & 16x16 PWR Lattice Type Core Designs ML20057B5431993-06-26026 June 1993 Errata for Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments, for Use in Issuance of Final Approved Version of Topical Rept ML20128B8101993-01-19019 January 1993 Safety Evaluation Accepting Methodology Described in Topical Rept RXE-91-002 Reactivity Anomaly Events Methodology for Reload Licensing Analyses for CPSES ML20126E0381992-12-0909 December 1992 Safety Evaluation Accepting Topical Rept NEDC-31753P W/Ter Recommendations W/Listed Exceptions ML20056D9351991-01-11011 January 1991 Topical Rept Evaluation Accepting Proposed Methodology for Fuel Channel Bowing Anaylses & for Referencing in Reload Licensing Applications W/Listed Conditions ML20235Q7121989-02-22022 February 1989 Safety Evaluation Re Review of WCAP-10271,Suppl 2 & WCAP-10271,Suppl 2,Rev 1 on Evaluation of Surveillance Frequencies & out-of-svc Times for ESFAS ML20206L9611988-11-23023 November 1988 Topical Rept Evaluation of PECO-FMS-0004, Methods for Performing BWR Sys Transient Analysis. Rept Approved,But Limited to Util Competence to Use Retran Computer Code for Facility ML20205M0091988-10-25025 October 1988 Safety Evaluation of Topical Rept YAEC-1300P, RELAP5YA: Computer Program for LWR Sys Thermal-Hydraulic Analysis. Program Acceptable as Licensing Method for Small Break LOCA Analysis Under Conditions Stipulated ML20204G8371988-10-18018 October 1988 Safety Evaluation Accepting Topical Rept 151, Haddem Neck Plant Non-LOCA Transient Analysis, Except for Issue of Feedwater Event ML20155G7991988-10-12012 October 1988 Topical Rept Evaluation of TR-045, BWR-2 Transient Analysis Using Retran Code. Methods Described in Rept Acceptable for Reload Analysis When Listed Conditions Satisfied ML20155G3201988-09-26026 September 1988 Safety Evaluation of TS NEDC-30936P, BWR Owners Group TSs Improvement Methodology. GE Analyses Demonstrated Acceptability of General Methodology for Considering TS Changes to ECCS Instrumentation Used in BWR Facilities ML20155B0501988-09-22022 September 1988 Topical Rept Evaluation of Suppl 1 to NEDC-30851P, Tech Spec Improvement Analysis for BWR Control Rod Block Instrumentation. Analyses Acceptable to Support Proposed Extensions to 3 Months ML20151K9921988-07-26026 July 1988 Topical Rept Evaluation of Nusco 140-1 Northeast Utils Thermal Hydraulic Model Qualification,Vol 1 (Retran). Rept May Be Generally Ref in Future Licensing Submittals.Further Justification by Util Required ML20150D9651988-03-21021 March 1988 Topical Rept Evaluation of Rev 0 to TR-040, Steady State & Quasi-Steady State Methods for Analyzing Accidents & Transients. Util Methods Acceptable for Performing Reload Assembly Mislocation Analysis W/Listed Exceptions ML20150D7671988-03-21021 March 1988 Topical Rept Evaluation of Rev 0 to TR-033, Methods for Generation of Core Genetics Data for RETRAN-02. Uncertainties in Input Parameters & Impact on Retran Results Should Be Determined for Qualification of Model ML20236D2621987-10-21021 October 1987 Topical Rept Evaluation of CEN-348(B)-P, Extended Statistical Combination of Uncertainties. Rept Acceptable ML20235D7041987-09-22022 September 1987 Safety Evaluation of Rev 0 to Topical Rept TR-021, Methods for Analysis of BWRs Steady State Physics. Rept,Methodology & Util Use of Methodology Acceptable ML20239A5461987-09-0909 September 1987 Safety Evaluation Supporting A-85-11, Retran Computer Code Reactor Sys Transient Analysis Model Qualification for Use in Performing plant-specific best-estimate Transient Analyses at Plant ML20215M3591987-05-0606 May 1987 Safety Evaluation Supporting Util Use of Suppl 1 to MSS-NA1-P, Qualification of Reactor Physics Methods for Application to PWRs of Middle South Utils Sys ML20212M7781987-02-17017 February 1987 Topical Rept Evaluation of WCAP-10325, Westinghouse LOCA Mass & Energy Release Model for Containment Design - Mar 1979 Version. Rept Acceptable for Ref in Licensing Actions ML20210N7331987-02-0404 February 1987 Safety Evaluation Supporting CEN-161(B)-P,Suppl 1-P, Improvements to Fuel Evaluation Model. Mods to Fission Gas Release & Fuel Thermal Expansion Models Acceptable ML20215B2331986-12-0404 December 1986 Corrected Page 1 to 861031 Topical Rept Evaluation of Rev 2 to STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Word Effective Inserted Before Words Pore Sizes in First Line of 4th Paragraph ML20214C5221986-11-14014 November 1986 Topical Rept Evaluation of Rev 0 to TR 020, Methods for Analysis of BWR Lattice Physics. Collision Probability Module Code Acceptable for BWR Fuel Lattice Calculations ML20213F6531986-11-10010 November 1986 Safety Evaluation of Rev 2 to Vol 3 of XN-NF-80-19(P), Exxon Nuclear Methodology for Bwrs,Thermex:Thermal Limits Methodology Summary Description. Rept Acceptable for Ref in Licensing Applications ML20207A8281986-11-0505 November 1986 Suppl 3 to Topical Rept Evaluation Re Submittal 2 to Rev 3 to CEN-152, C-E Emergency Procedure Guidelines. Rept Acceptable for Ref ML20215N6901986-11-0404 November 1986 Topical Rept Evaluation of BAW-10155, FOAM2 - Computer Program to Calculate Core Swell Level & Mass Flow Rate During Small-Break Loca. Rept Acceptable W/Listed Restrictions Re Ranges of Core Flow Rate & Pressure ML20215N3921986-10-31031 October 1986 Topical Rept Evaluation of STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Rept Acceptable for Ref in License Applications ML20211D6921986-10-16016 October 1986 Safety Evaluation of Nusco 140-2, Nusco Thermal Hydraulic Model Qualification,Vol II (Vipre). Rept Acceptable for Establishing Input Values & Selection of Correlation Options & Solution Techniques for Calculations ML20206S6511986-09-15015 September 1986 Topical Rept Evaluation of Addenda 3 to WCAP-8720, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations/Application for BWR Fuel Analysis. Rept Acceptable for Ref in Licensing Applications ML20212N2001986-07-23023 July 1986 Topical Rept Evaluation of Rev 1 to XN-NF-85-67 (P), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel. Rept Acceptable as Ref for Application to Jet Pump BWR Reload Cores,W/Listed Conditions 1994-08-25
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pEc 111985 EVALUATION OF EXXON TOPICAL REPORT XN-NF-825(P)
(TACS 57460)
By letter dated April 8, 1985, Exxon Nuclear Company (ENC) submitted a proprietary Licensing Topical Report (LTR) XN-NF-825, "8WR/6 Generic Rod ;
Withdrawal Analysis, MCPRp", for staff review. The report describes the ENC analysis of the rod withdrawal event (RWE) for reactors (such as the BWR/6) which contain a Rod Withdrawal Limiter to limit the consequences of the "at power" rod withdrawal. We have reviewed the document and prepared the following evaluation.
- 1. Description of Report After an introduction (Section 1) and summary (Section 2) the report presents a brief overview (Section 3) of the protection systems for the rod withdrawal transient. The Rod Block Monitor (BWR/2-5) and the Rod Withdrawal Limiter (BWR/6) are described. It is noted that the analysis of core behavior during a rod withdrawal event is the same for both protection systems. The Technical Specifications for use with the Rod Withdrawal Limiter are described and typical curves are presented.
Section 4 of the report describes the methodology used to perform the statistical analysis required to demonstrate at a 95% percent confidence level that MCPR safety limits are not violated for at least 95 percent of the rod withdrawal events. The choice of initial conditions for the event simulations is described and the methods and procedures used to perform the simulations are discussed.
Section 5 contains the results of the analyses. Section 6 presents the statistical methodology employed. The application of the generic analyses to reload support is discussed in Section 7. _,.-
8602280650 DR 860214 TOPRP ENVEXXN PDR ENCLOSURE
- 2. Description of Evaluation The Exxon generic analysis of the rod withdrawal event assumes the presence of the Technical Specification on rod withdrawal limits as a function of power which have been established by the reactor designer (General Electric). These limits are one foot for core powers greater than 70 percent of rated power and .
two feet for core powers between 20 and 70 percent of rated. The analyses are then performed to establish the values of operating limit MCPR as a function of core power which are required to assure that the Safety Limit MCPR will not be violated for the rod withdrawal event. This is an acceptable approach.
The calculational methods and procedures used to simulate the rod withdrawal event are essentially the same as those described in Reference 1 which has been approved by the staff (Reference 2). The only difference from the methods of Reference 1 occurs because of the different means of effecting the rod block. This block is triggered by local power rise in the analyses of Reference 1 and by a fixed rod withdrawal distance in the BWR/6 analysis. This difference does not affect the calculation of the change in MCPR as a function of rod withdrawal. We thus conclude that the methods used to simulate the rod withdrawal event are acceptable.
A data base of initial conditions was constructed by choosing actual operating states obtained for a BWR/6 reactor. These states were chosen so that the entire power-flow map of the reactor was represented and the range of the
_ control rod patterns throughout the cycle was included. Emphasis was placed on states that are more of ten represented (e.g. , those new full flow-full power) and 37 statepoints were chosen.
The next step in the process consists of simulating the withdrawal of each gang of control rods in the data base in two one-foot incret.ents and obtaining the resulting Critical Power Ratio for each fuel assembly in the core at each increment. Calculations were performed to show that ganged withdrawals are always more limiting than single rod withdrawals. This result is consistent with previous analyses and is acceptable. .
-y
The core AMCPR values for each foot of control rod withdrawal are obtained by taking the difference between the initial and final state core MCPR values. A .
statistical analysis is then performed to determine the 95/95 generic rod withdrawal error transient limits. These calculations were performed with the Exxon XTGBWR code which has been approved (Reference 2) for BWR. core simulation. . ,
The statistical evaluation of the 95/95 value of the expected initial MCPR values is straightforward. The values of the mean expected MCPR as a function of power are well above the typical Technical Specificaticn limit for this quantity. Adherence to this limit is not expected to result in operational inconvenience.
The statistical analysis of the AMCPR data is complicated by the presence of zero values among the data base which invalidate the assumption of normality of the random variable. The pr'oblem was resolved by regarding AMCPR=0 as a censoring point with positive values of AMCPR representing that portion of the random distribution which is observable. This is an acceptable treatment of the statistics.
The operating limit MCPR required by the Rod Withdrawai Error transient may be obtained by adding the AMCPR value as a function of power to the safety limit value. This results in a value of 1.13 at 100 percent power, for example.
The typical Technical Specification value at 100 percent power is 1.18-1.20.
Similar margin exists at other powers.
The values of MCPR operating limit (OLMCPR) as a function of power are typically determined by other events (e.g., LOCA and flow-run out events).
Section 7 of the present topical report presents a curve of OLMCPR as a function of power which is required for the RWE. This represents a lower f limit for this quantity. As noted above considerable margin exists between this curve and the current Technical Specification curve.
The results described above are valid for operation within the power flow, domain illustrated in Figure 4.1 of the present report and for theifuel management scheme used for determining the operating states of the data base.
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Use of other power flow domains (e.g. , the Maximum Extended Operating Domain-ME00) or other fuel management schemes (e.g., the single sequence loading pattern) will require verification by analysis that the conclusions of this report are valid.
- 3. Regulatory Position -
Based on our review which is described above we conclude that Exxon Topical Report XN-NF-825 may be used as a reference to the analysis of the Rod Withdrawal Error transient for BWR/6 reloads. We further conclude that a (
cycle specific analysis of this event is not required provided that the operating power-flow region is bounded by that used in the analysis reported in the present topical report and the core loading and control rod patterns are consistent with those used in the data base. Reloads which do not meet these provisions will need analyses to confirm the conclusions reached in the present topical report.
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6 REFERENCES l
- 1. XN-NF-20-19(P), Including Supplements 1 and 2, " Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, May, 1981.
- 2. Letter to G. F. Owsley, Exxon Nuclear Company, from J. R. Miller, NRC dated April 7, 1982.
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