Topical Rept Evaluation of Suppls 1 & 2 to XN-NF-507(P), Exxon Nuclear Co Setpoint Methodology for C-E Reactors- Statistical Setpoint Methodology. Rept Acceptable for Referencing in License ApplicationsML20138P921 |
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Office of Nuclear Reactor Regulation |
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ML20138P916 |
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NUDOCS 8512260350 |
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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
[Table view]Some use of "" in your query was not closed by a matching "". |
Text
. . _ _ - .- _ _ _ - . . . - - . - -- ~ _ . - - .
. i j
I Enclosure Evaluation of the Exxon Statistical Setpoint
- Methodology for CE Reactors l
- 1. Indroduction ,
i Exxon Nuclear Company (ENC) has developed methods for statistically combining the parameter and measurement uncertainties that are applicable to themal , ,
margin analysis for CE reactor reload cores. This ENC methodology as documented l . in Referencer 1, 2 and 3 is usedito replace or validate determir..stic reactor l local power density (LPD) and thermal margin / low pressure (TM/LP) trip setpoints r ^and limiting conditions for operation (LCO) with 95% probability bounds (with .
95% confidence). ,
We have reviewed the ENC statistical thermal margin setpoint analysis method-
['
ology for CE reactors. The review was based on the ENC reports pertaining to
., .their setpoint methodology (Refs. I to 5), and on the ENC responses (Ref. 8) to the review questions. The results of the review are discussed in this report as f follows. ,
[ , .i
.2. The Evaluation of Methodology for the LPD Trip Setpoint and LCO k
The LPD trip setpoint is established to scram the reactor before the peak fuel ,
( centerline temperature reaches the melting temperature, while the LPD LCO is an ;
. operational limit on reactor power to assure that the calculated peak cladding
. temperature will not exceed the safety limit of 2200*F during a large break
. loss-of-coolant-accident. ,
f In the ENC LPD setpoint calculations, the radial power peaking factor with the worst l
. axial' power distribution which results in the highest peaking for LPD thermal margin is used for conservatism.
l t 0512260350 851216 '
PDR TOPRP ENVEXXN C PDR ,
I y.;. - =. a. . - -.. a . . .. . - _
. As indicated in the topical report, the power measurement and power peaking uncertainties are related to the nominal power trip setpoint (calculated based
, ontheinputparametersatbestestimatevalues)toobtainanallowablepower l trip setpoint. TheMonteCarloSimulation(MCS)isusedtostatistically combine the power peaking and fuel rod engineering factor uncertainties. The I convolution integral method is used to combine the uncertaintie's for the power e
' measurement and the power trip signal process. We have reviewed the ENC u~se
- of MCS and the convolution integral method. We find that MCS is a valid g
technique for combining the probability densities of two random variables and
' the equations used in the convolution method are correct. Since the highest peaking, which results in a lower LPD thermal margin, is used and the methods
< to combine the uncertainties involved in the LPD setpoint are correctly used,
[ -
we conclude that the methods for determination of LPD trip setpoint and LCO are acceptable. (The results of review for MCS are further discussed in Section 6).
- 3. The Evaluation of TM/LP Trip Setpoint Methodology I
! The TM/LP trip is established to scram the reactor to prevent the DNBR in the limiting coolant channel in the core from decreasing below that value
- corresponding to a 951 probability that DNB will not occur, at 95% confidence t level.
~
i 'An application technique common to both the TM/LP trip setpoint analysis and l the DNB thermal margin allowable power LCO (discuss'ed in Section 4) is the use F of a "most sensitive" ASI point for the statistical analysis. The "most
) : sensitive" paint is the value of ASI, and its assocf ate'd power distributions
- - - and operating conditions, where the largest difference between a deterministic 7 setpoint (calculated with all key parameter uncertainties set to 95% probability '
'I, or the tolerance bounds) and a nominal setpoint (calcul'ated with all key parameters set at nominal values) is calculated. The resulting 95% probability i bound ~at a 951 confidence (the 95/95 bound) for the trip setpoint or LCO
' ~ calculated at this "most sensitive" ASI point is th'en applied to the calculated L
b nominal values for the setpoint at all other ASI points.
i 2
- y. . _ . . . . _ . ._ _.
_ . . _ . . . .. j
t We have reviewed the TM/LP trip setpoint methodology and found that the methodology I for defining the 95/95 bound for the TM/LP setpoint'is an adequate approach; the p methodology for verification of the existing TM/LP setpoint algorithm'is a contervative application of the statistical methodology. Therefore, we conclude that the TM/LP trip setpoint methodology is acceptable.
- 4. The Evaluation of DNB Allowable Power LCO Methodology ,
'i The DNB allowable power LCO (DNB LCO) is an operational limit on power as a
,. function of ASI, designed to protect the fuel against DNB (i.e., a 955 probability with a 95% confidence that DNB will not occur on the hot fuel red) during any ,
limiti~ng Anticipated Operational Occurrence (A00). The DNB LCO is based on the k- --results of thermal-hydraulic analyses for limiting A00s. Potentially limiting g A00s were considered to define which transients are limiting for the DNB LCO.
For CE reactors with an asymmetric steam generator protection' trip,'th'e identified limiting A00s for the DNB LCO are the loss of flow'(due to loss of L . pumping power) or the control element assembly drop' transients. Both of these transients were analyzed to define which was most limiting for the DNB LCO.
f g
Thermal-hydrualic calcualtions are done using the XCOBRA-IIIC computer code (Ref. 11) and the XNB DNB correlation (Ref. 10), which were previously approved F .byNRC(Ref.6),todeterminethereactorpowerforwhich'thetransientminimum
[ : DNBR reaches the limit for DNBR as a function of key statistical variables. ,
A response surface analysis performed at the "most sensitive" ASI point (defined'
! inSection3)isusedtoformulate,fromXCOBRA-IIICcalcualtions,anequation
~
I (the response surface equation) for the power limit (for DNBR = DNBR safety L
limit),asafunctionofthestatisticalparameters. Monte Carlo Simulation s- (MCS) is used to propagate the uncertainties in the statistical parameters '
- '. through the response surface equation. The MCS results are then used to define ,
! a 95/95 bound for the DNB allowable power LCO.
I i
i~ 3 u 1:..-- =-~__ = _ -~
! We have reviewed the statistical methodology used for the DNB LCO analysis .
p and found that an approved computer code and DNB correlation are used and, the response surface analysis and Monte Carlo Simulation are adequate methods for calculating the uncertainties of the statistical parameters. We conclude ,
that the DNB LC0 methodology is acceptable. -
Y ,
- 5. The Evaluation of Application of the XNB DNB Correlation The ENC analyses for establishing the 95/95 bound for the TM/LP trip setpoint I and the DNB LCO involve calculating the margin from an approach to DNB by use of the XNB correlation (Ref. 10). The XNB correlation and its safety limit l 1
DNBRof1.17werepreviouslyapprovedbyNRC(Ref.6). However, the XNB correlation I
was based on DNB data for many different PWR fuel bundle geometries, and is p . thus a general PWR CHF correlation rather than a correlation defined for a -
}.
specific fuel bundle design. As restricted by NRC in Reference 6, when the XNB p correlation is applied for a specific type of fuel bundle which does not belong
- - to the data base on which the XNB correlation was based. ENC should submit for NRC approval a compensating fact;r to account for the effect of the f - different fuel type on the DNBR calculation. -
- 6. The Evaluation of Monte Carlo Simulation Methodology i
F* In Monte Carlo Simulation, a random sampling routine is used to choose values - .
! for the input statistical variables according to their defined probability /
distributions, and then to exercise the relevant computer program. The MCS ;
routine will continue this process for a specified number of trials, after '
fa which many output points are available for defining the probability distribution k of the response variable by use of a histogram, and for defining the statistical
! characteristics of the response variable.
i , i TheM0CARScomputercode(Ref.12)wasusedtoperformMonteCarloSimulation
('- for the ENC Statistical Setpoint analyses. In M0 CARS, values for each input a variable are randomly chosen from the uncertainty distribution of the respective !
l variables.
i 4
, .= . -.. _. = _ . . = - _ . - . ..
a
- i. .
.We have reviewed the ENC use of MCS. We find that ENC sets the accuracy high i enough(toinclude99.5%ofthepopulationwitha99.5% confidence)to' .
adequately envelope the 95% probability bound. Based on the review of 1 . Reference 13, we find that 95/95 bound is conservatively defined using ,
distribution free order statistics, which bounds the true value for the desired probability level. Therefore, we conclude that the ENC use of MCS_is adequate
'and acceptable.
- 7. The Evaluation of Responses Surface Analysis Methodology In a response surface analysis (RSA), the computer models which are used to a ' calculate the value of the response variable, such as DNBR,'re a exercised many .
times with different values for the key input variables in order to define a I response surface; that is, the values of the response variables as function of
. . the key input variables. The response surface is fitted by a response surface equation (RSE), a polynomial, by using a regression analysis. The RSE can be used as a substitute for the more complex algorithms or computer model.s for the !
many calculation trials needed to define the probability distribution of the response variable. ENC uses MCS approach to define the.95/95 probability bound in the RSA application. '
The ENC use of RSA in their statistical setpoint methodology as described in, b their generic statistical uncertainty analysis methodology report was previously ;
p ,
reviewedandapprovedbyNRC(Ref.7). We also find that all the limitations
( ; (Ref. 7) imposed by NRC on the ENC RSA are adequately met. ;Therefore, we ,
conclude that the ENC use of RSA methods is acceptable.
t I' 8. Review of Input Parameter Uncertainties 1
- 8.1 The Staff Evaluation of Parameters Uncertainties i
b'4 . ENC generally assumes for non-data based parameter uncertainties a normal dis-tribution.. However, numerous examples of non-normal distributions for parameter
'r uncertainties were found. Use of a normal distribution to represent a uniform distribution was also found to be not neccessarily conservative (Ref. 9).
5 -
N l
L We find that the general ENC approach for the non-data based uncertainties is -
- l. not adequately supported, since the distribution of the uncertainties for plant
' system parameters may be a function of how the measurements are taken and of i- how the sigr.als are processed. The magnitudes of the measurement uncertainties for system parameters as listed in References 2 and 8 appear to be conservative e when compared to data for measurement uncertainties. However, ENC has not adequately supported the appropriateness (or conservativeness) of their assumption of normal distribution for all measurement uncertainties. Therefore, we require ENC to justify the measurement uncertainties and the assumed probability distributions for specific CE reactor core reload application (once for each -CE
, ' reactor), or provide bounding values for all CE reactor reloads (once for all f CE reactors). The affected system parameters are. reactor power, ASI, pressure, inlet coolant temperature, coolant flow, and scram delay time.
r 8.2 The Staff Evaluation of Power Distribution Uncertainty
' ENC used the inferred power distribution and various PWR power distribution data sources to determine the power distribution uncertainty. However, as is discussed below, the PWR power distribution data used by ENC may not be p applicable to CE reactor cores.
ENC indicates that the measurement data base for the inferred detector power' ;
uncertainty is derived from measurements of the detector signals taken during
. Cycles 4 and 5 at the St. Lucie Unit 1 plant. However, the measurement data base for axial power distribution were derived from the measurements performed at two Westinghouse reactors. Also, the measurement data base for the uncertainty in the assembly local peaking factor is composed of measurements performed by Babcock & Wilcox and Battelle Pacific Northwest Laboratories.
The above data bases for Westinghouse and Babcock & Wilcox fuel may not be applica-
{
ble to CE reactors because of potential significant neutronic differences between CE fuel bundles and cores ar.d the other PWR fuel bundles and cores.
f Therefore, we require that the data or analyses used to establish or, justify
! the power distribution specifically address the CE reactor core and CE fuel bundle design, i
I 6
0 . . . . .
e c .
i 8.3 The Staff Review of Operating Point Uncertainty In the statistical setpoint analysis, ENC uses deterministic (worst case) assumptions to specify all operating state variables at the technical specifi-cationboundingvalues(oroperationalbounds)whichminimizetheDNBthermal f margin. We find that this approach is conservative and is acceptable.
- 9. Conclusions
' Based on the review described above we conclude that the topical reports XN-NF-507 (Supplements 1 and 2) are acceptable for referencing in licensing I actions by ENC with respect to the Exxon statistical setpoint methodology for 5 'CE reactors. The following restrictions apply to the use of the methodlogy:
- 1. The statistical setpoint methodology, as defined by References 1, 2, 8, and in this report should not be altered or modified, except as approved t herein, without additional NRC review.
b 2. ENC is required to justify the system parameter measurement uncertainties E and their probability distributions for specific CE reactor core reload I
applications of the methodology. The affected system parameters are:
reactor power, ASI, pressure, inlet coolant temperature, coolant flow and I ' scram delay time. The justification should include consideration of the
- measurement system and signal processing when defining a'n appropriate probability density function for the measurement uncertainty.
i k i P .3. ENC is required to justify the power distribution uncertainty assumed for
- - specific applications of the methodology for CE reactor core reloads.
' The justification will need to demonstrate that the power distribution 4 < uncertainty is applicable and/or conservative based on analyses and/or m data applicable to CE fuel and cores.
- 4. ENC should submit for NRC review and approval a CHF uncertainty compensating
- factor to include an adequate allowance for the uncertainty in the CHF l
- when applied to fuel bundle designs different from test bundles used for obtaining the DNB test data, i
7
[
th ~ . . ~ _ . . . . . . . . . - - - , . - - - - - - .
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3 , , . __ _ _ ____ _
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