ML20151V284

Nonproprietary Extended Statistical Combination of Uncertainties

ML20151V284
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Site:	Calvert Cliffs
Issue date:	01/01/1988
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CEN-348(B)-NP-A, NUDOCS 8808220206
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LEGAL NOTICE THIS REPORT WAS' PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COM9USTION 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 MERCHANTABILITY, WITH 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 LIABILITIES WITH RESPEt,T YO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.

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CEN-348( B )-NP-A W

4 Exterided Statistical Combination of Uncertainties e

Combustion Engineering, Inc.

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[%, I y UNITED STATds 7,

3 NUCLEAR REGULATORY COMMISSION WASN188010N, D. C. 206H j

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October 21, 1987 Docket No. 50-317 i and 50-318 1 mr. J.A. Tiernan Vice President-Nuclear Energy Baltimore Gas and Electric Company P.O. Box 1475 '

i Baltimore, Maryland 21203

SUBJECT:

SAFETY EVALUATION OF TOPICAL REPORT CEN-348(8)-P, "EXTENDED l

STATISTICAL COM8INATION OF UNCERTAINTIES" (TACS 6498j We have completed our review of your letter dated March 17, 1987 which submitted the Combustion Engineering (C-E) Topical Report CEN-348(3) P "Extended Statistical Combination of Uncertainties," (ESCU) for Commission i

review and approval. I This report provided a methodology for statistict.11y combining the uncertainties involved in the departure from nucleate boiling related analog protection and monitoring system setpoints.  !

The Comission finds that the use of this ESCU methodology is acceptable for-C-E 14x14 fuel assemblies at Calvert Cliffs Units I and f. '

Enclosed herewith is our Safety Evat etton.

Sincerely, '

Scett Alexander McNeil, Project Manager Project Directorate I-1 Division of Reactor Projects I/II Enclosuer:

As stated cc: See next page N

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Mr. J. A. Tiernan Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant ec:

Mr. John M. Gott, President Regional Administrator, Region !

Calvert County Board of U.S. Nuclear Regulatory Comission Comissioners Office of Executive Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue D. A. Brune, Esq. King of Prussia. Pennsylvania 19406 G2neral Counsel .

Baltimore Gas and Electric Company P. O. Box 1475

- Baltimore, Maryland 21203 Jay E. Silberg, Esq.

Shaw, Pittman, Potts d Trowbridge 1800 M Street, NW Washington, DC 20036 Mr. M. E. Bowman, General Supervisor Technical Services Engineering Calvert Cliffs Nuclear Power Plant MD Rtc 2 & 4, P. O. Box 1535 Lusby, Maryland 20657 0073 Resident inspector c/o U.S. Nuclear Regulatory Comission ',

P. O. Box 437 Lusby, Maryland 20657-0073 Bechtel Power Corporation .

ATTN: Mr. D. E. Stewart Calvert Cliffs Project Engineer 15740 Shady Grove Road ,

Gaithersburg, Maryland 20760 Combustion Engineering, Inc.

ATTN: Mr. W. R. Horlacher, III Project Manager

. P. O. Box 500 1000 Prospect Hill Road Windsor, Connecticut 06095-0500 l Department of Natural Resources Energy Administration, Power Plant Siting Program ATTH: Mr. T. Magette Tawes State Office Butiding Annapolis, Maryland 21204 1

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[** "% UNITED sT ATEs NUC!, EAR REGULATORY COMMISSION 3

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, SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

. BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS. 1 AND 2 e l

. DOCXET N05. 50-317 AND 50-318

- REVIEW OF TOPICAL REPORT CEN-348(B)-P j "EXTENDED STATISTICAL COMBINATION OF UNCERTAINTIES"  !

. l l

1. INTRODUCTION By letter dated March 17, 1987, Baltimore Gas and Electric Company (BGAE) {

requested NRC review and approval of CEN-348(8)-P. "Extended Statistical Combination of Uncertainties". The extended statistical combination of ,

uncertainties (ESCU) method is an enhancement of the existing statistical '

combination of uncertainties (SCU) methodology previously reviewed and approved I by the NRC (Ref. 1). The report describes an improved method for statistically '

combining the uncertainties involved in the departure from nucleate boiling .

4 (DNB) related analog protection and monitoring system setpoints. - 1 The licensee has defined the input data required for a detailed thennal- '

hydraulic analysis by type: (1) system parameters which describe the physical I system and are not monitored during reactor operation and (2) state parameters -

• which describe the operational state of the reactor and are monitored during i t operation. There is a degree of uncertainty in the value used for each of the input para:neters used in the design safety analyses. In the past, these uncertainties had been handled by assuming that each variable affecting DN8 was at the most adverse limit of its uncertainty range. The assumption that all factors are simultaneously at their most adverse values leads to conservative restrictions in reactor operation. Therefore, the SCU methodology wu developed

. to statistically combine uncertainties in the calculation of new limits for

,Calvert Cliffs; These limits ensure with at least a 95 percent probability and a 95 percent confidence level (95/95) that neither DNB nor fuel centerline melt will occur. Part 1 of the methodology (Ref. 2) described the application of (LPD) and thermal margin /

the SCU to the development of the local power density (LSSS).

lowpressurer(TM/LP)limitingsafetysystemsettings These are used in the analog reactor protection system to protect ag' Inst fuel centerline melt and DNB, respectively. Part 2 (Ref. 3) used SCU methods to develop a new i DNB catio (DNBR) limit. Part 3 (Ref. 4) used SCU methods to define limiting l conditions for operation (LCO).  !

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i 2.0 EVALUATION As mentioned tainties above,The in two groups. the uncertainties existing SCUinmethod (Refs.

one group system 2, 3.(and prameter 4) treats uncer-

. uncertainties and critical heat function (CHF) correlation urcertainties) are statistically combined to generate a DNSR probability density function. The 95/95 probability / confidence level limit of this function is then used as the

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setpoint analysis minimum DNBR. The uncertainties in the other group (state

' parameter uncertainties., axial shape index ( ASI) uncertainties, and processing uncertainties) are statistically combined to generate overpower error prob-

- ability density functions for LSSS ano LC0 processes. The 95/95 limits of these functions are then applied as overpower penalties in the generation of LS$$ and LC0 setpoints.-

Although the uncertainties within each group are combined statistically and a 95/95 probability / confidence level generated for each group, the resultant '

uncertainties of the two groups are effectively combined in a deteministic manner due to the separate application of the two uncertainty limits. The proposed ESCU methodology would incorporate the DN8R probability density function, which is generated by statistically combining the system parameters i and CHF correlation uncertainties, into the protection and monitoring system stochastic simulation models together with the ASI, state parameter, and processing uncertainties.

The staff has reviewed the uncertainties and the uncertainty treatmeh; procedure described for the proposed ESCU methodology and has determined that the resultant penalties applied to the setpoint calculations adequately incorporate all un-certainties at the 95/95 probability /cenfidence level. Tte analytical methods ,

reviewed and approved show' that for CE 14x14 fuel, a DNBR limit of 1.15 with the uncertainty penalties derived in the report provides a 95/95 probability / confidence ,

1evel assurance against DNB occurring during steady state operation or anticipated operational occurrences.

3.0 CONCLUSION

The staff has reviewed the ESCU mtthodology presented in CEN-348(8)-P and finds it to be an acceptable method for statistically combining uncertainties for the TM/LP LSSS and DNB LCOs for Calvert Cliffs Units 1 and 2 utilizing CE 14x14 fuel.

4.0 REFERENCES

1. LetterfromD.H.Jaffe(NRC)toA.E.Lundvall(8G4E),"RegardingUnit1 Cycle 6 License Approval (Amendment #71 to OPR-53 and SER). Appendix A to Attachment, June 24, 1982.

2. "Statistical Combination of Uncertainties Part 1," CEN-124(B)-P December 1979.

3. "Statistical Combir,ation of Uncertainties Part 2." CEN-124(B)-P January 1980.

4 "Statistical Combination of Uncertainties Part 3." CEN-124(B)-P, March 1980.

ABSTRACT The Extended Statistical Combination of Uncertainties (ESCU) report describes an improved method for statistically combining uncertainties for the C-E calculated Thermal Margin / Low Pressure (TM/LP) LSSS and Departure from Nucleate Boiling (DNB) Limiting Conditions for Operation (LCO) for Calvert Cliffs Units I and II. ESCU is a modification to the NRC approved statistical Combination of Uncertainties (SCU) method currently in use for Calvert Cliffs plants.

This report cascribes the ESCU method of calculating total uncertainties expressed in ,

ssure with at least a 95% probability at a 95% confidence level that the hot:est fuel rod in the core does not experience a Departure from Nucleate Boiling (DNB) during normal operation or an Anticipated Operational Occurrence (A00) initiated within the LCO limiti.

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i TABLE OF CONTENTS CHAPTER PAGE 1.0 Introduction

. 1.1 Purpose 1-1 1.2 Background 1-1 1.2.1 Protection and Monitoring System 1-1 1.2.2 Uncertainties and Uncertainty Treatment 1-1 Procedure 1.2.3 Application of Uncertainties 1-1

,- 1.3 ESCU Concept 1-1  :

1.4 Report Scope 1-9 1.5 Sumary of Results 1-9 1.6 References for Section 1.0 1-9 2.0 Analysis 2.1 Objectives of Analysis 2-1 2.2 Analytical Techniques 2-1 .

2.3 TM/LP LSSS Stochastic Simulation 2-1  !

2.4 DNB LCO Stochastic Simulation 2-1  !

2.5 References for Section 2.0 2-2 (

1.0 Results and Conclusions 3.1 Results of Analysis 3-1 3.2 Conclusion 3-1 3.3 References for Section 3.0 3-1 Appendix i A.0 MONBR Probability Censity Function I

, A.1 Background A-1 A.2 MDNBR PDF for Extended SCU Analysis A-1 A.3 References for Appendix A A-6 B.O Proposed Changes to Related Technical Specificatiens Bases B.1 Discussion B-1 j l

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LIST OF TABLES TABLE PAGE 1-1 Uncertainties Included in the Original SCU 1-3

, ONBR Probability Density Function  ;

1-2 Uncertainties Treated in the Original 1-4 Protection (TM/LP LSSS) and Monitoring (DN8 LCO) Systems Simulation Models 3-1 Si;atistically Combi.ned Uncertainties for 3-2 a

A-1 Component System Parameter Uncitrtainties A-3 .

and Allo,vances Accommodated by  :

pod f. -

B-1 Technical Specifications Bases Requiring B-2 Change Oue to Incorporation of the ESCU l

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LIST OF FIGURES FIGURE ,

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1-1 SCU Thermal Margin U1 certainty Analysis 1-5

- 1-2 SCU Ex-Core Detector Monitored DNG LCO 1-6 Uncertainty Analysis 1-3 SCU Approach 1 1-4 ESCU Approach 1-8 t

2-1 (ESCU) Thermal Margin Uncertainty Analysis 2-3 2-2 (ESCU) Ex-Core Detector Monitored DNB LCO 2 Uncertainty Analysis

, A-1 DNBR Probability Distribution Function A-4 A-2 Channel Numbering Scheme for Stage 1 TORC A-5 Analysis to Establish Response Surface State 1 Parameters o

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4 DEFINITION OF ACRONYMS AND ABBREVIATIONS A00 Anticipated Operational Occurrence (s)

BASSS Better Axial Shape Selection System

• 1 B,g Available Overpower Margin SMU Power measurement uncertainty CETOP Computer code used to deternine the overpower limits due to thennal-hydraulic conditions CHF Critical Heat Flux DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Soiling Ratio ESCU Extended Statistical Combination of Uncertainties F Primary Coclant flow rate F

r Integrated Radial Peaking Factor GWO/MTU Giga Watt Day per Metric Ton Uranium .

I Core average axial shape index 1 I p

Peripheral axial shape index LC0 Limiting Condition (s) for Operation LSSS Limiting Safety System Setting (s)

MONBR Minimum DNBR I

', 0.0. Outside Diameter P Reactor coolant System Pressure l j- p.d.f. pecbability density function Pfdn Power to DNBR SAFDL SAFDL Specified Acceptable Fuel Design Limit (s)

SCU Statistical Combination of Uncertainties T Reactor coolant cold leg temperature T

h Reactor cooiant hot leg temperature TM/LP Thermal Margin / Low Pressure ,

TCRC Thennal Hyd"aulic Calculation Medel y

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1.0 INTROCUCTION 1.1 Purpose The purpose of this report is to describe an improved method for statistically combining the uncertainties involved in the Departure from Nucleate Boiling (DNB) related analog protection and monitoring system setpoints. This improved method, Extendec, Statistical CombinationofUncertainties(ESCU),consistsof, The re.sult is an overall uncertainty penalty expre,ssed in terms of which are applied in DNB LSSS and LCO calculations.

1.2 Background

1.2.1 Protection an'd Monitoring System The basic purposes and interactions of the ONB LSSS and LCO were previously described in Section 1.2.1 of Reference 1-1 and are applicable to this report.

1.2.2 Uncertainties and Uncertainty Treatment Procedure The uncertainties which have been considered and incorporated in the current SCU methodology have been described in Reference 1-1 (Appendices A1, A2 and A3), Reference 1-2 (Section 3) and 0.eference 1-3 (Appendix A). These uncertainties were treated in two groups.

One group of uncertainties, referred to as system parameter uncertainties, included engineering factors and other uncertainties used in the development of the TORC model. These uncertainties were statistically combined with the CHF correlation uncertainties (Ref.

1-2). The penalty associated with these uncertainties was expressed in terms of a higher DNB SAFDL. The second group of uncertainties, consisting of axial shape index uncertainties, measurement uncertainties (of state parameters), and processing  ;

uncertainties were combined through stochastic simulation models l representing CNB LSSS (Reference 1-1) and DNB LCO (Reference 1-3) calculation processes. These simulation models produced the penalty in terms of percent overpower.

1.2.3 Application of Uncertainties The methods used in application of the resulting SCU uncertainty penalties to the subject limits are presented in Reference 1-4.

l 1.3 ESCU Concept The SCU method mentioned in Section 1.2.2 and describe,d extensively l in References 1-1 through 1-3. treats uncertainties in a

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. .Therefore, even though the uncertainties within each part were

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for each groun, the resultant uncertainties of tie two

, groups we-o effectively . - due to Tablea 1-1 and "1-2 list the uncertainties treated by SCU method in DNBR probability density function and simulation models, respectively. These uncertainties are described in references 1-2, 1-1 and 1-3. Figures 1-1 and 1-2 show the original SCU DNB LSSS (TM/LP) and DN8 LCO simulation models. Figure 1-3 is a graphical representation of the original SCU method.

The Extended Statistical Combination of Uncertainties (ESCU) incorporates ,

(Ref. 1-2) into the protection and monitoring system stochastic <

simulation models which previously represented The resultant probability censity functions represent all uncertairities, anct the 95/95 probability / confidence litnits are applied as calculations. ' Figtve 1-4 is a graphical representation of the ESC <J method.

The motivation for introducing the ESCU method is to improve the.

plant operating space and flexibility by reducing the overly conservative uncertainty penalty allowances currently applied to the DNB LSSS and LCO limits in their setpoint calculations. The

. reduction in overall uncertainty penalties results from

instead of using the ori,ginal SCU approach which treated them

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TABLE 1-1 Uncertainties Included in the Original SCU DNBR Probability Density Function ,

, Core inlet flow distribution j Engineering factor on enthalpy rise

'- Systematic fuel rod pitch

. Systematic fuel clad 0.0. )

i Engineering factor on heat flux . l l

CE-1 Critical Heat Flux (CHF) correlation Fuel rod bow l 1

Note: For a complete description of these uncertainties, see I Section 3 of Reference 1-2.

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TABLE 1-2 Uncertainties Treated in the Original SCU Protection (TM/LP LS35) and Monitoring (DNB LCO) Systems Simulation Models Core power Primary coolant mass flow Primary coolant pressure Core coolant inlet temperature Power distribution (peaking factor) -

Axial shape index Note - For a complete description of these uncertainties, see Appendix A of References 1-1 and 1-3 P

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)AS & ELECTRIC CO. SCU Calvert Cliffs THERMAL MARGIN UNCERTAINTY ANALYSIS 1-1 Nweleer Power Pla,nt 1-5

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Calvert Cliffs i Nuclear Power Plant .

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ESCU APPROACH Calvert Cliffs 14 Nwclear Power Plant 14 4

1.4 Report Scope The objective of this report is to define the methods used to statistically combine uncertainties applicable to the Thermal Margin / Low Pressure (TM/LP) Limiting Safety System Settings (LSSS)-and the Departure from Nucleate Boiling (DNB) Limiting Conditions for Operation (LCO). The

, report encompasses the following issues;

1. To define the new DNB LSS$ and CNB LC0' stochastic simulation models which calculate the overall uncertainty penalties.

2. To define the which is input to the stochastic simulation models. <

3. To enluate the aggregate uncertainties and the as they are applied in the determination c'f the TM/LP LSS5 and DNB UC0.

1.5 Sumary of Results The analytical methods presented in section 2 are used to show that< the application of the ESCU method with an , ,

results 91 typical uncertainty penal, tits of for TM/1.P LSSS and !for ex-core and in-com DNB LCO.

1.6 References for Section 1:

1-1* "Statistical Combination of tincertainties dart 1" " N-124(B)-P December, 1979 1-2* "Statistical Combination of Uncertainties Part 2", CEN-124(B)-P January, 1980 1-3* "Statistical Ccmbination of Uncertainties part 3", CEN-124(B)-P March, 1980 1-4 "CE Setpoint Methodology", CENPD-199-P Rev. 1-P-A, January,1986 1-5 Letter. 0.H. Jaffe (NRC) to A.E. Lundvall, Jr. (BG&E), "Regarding Unit 1 Cycle 6 License approval,(amendment *71 to OPR-53 and SER)",

- June 24, 1982 These keferencet have been approved for use by the NRC in Appendix A of the Attachtrent to Reference 1-5.

1-9 t

2.0 ANALYSIS 2.1 Objectives of Analysis The objectives of the statistical analysis are to determine overall uncertainty factors to be applied to the TM/LP LSSS and the DNB LCO.

These uncertainty factors are determined such that there is a 95%

probability at a 955 confidence level that the combined effect of uncertainties on the LSSS and LCO limits will not exceed these factors.

2.2 Analytical Techniques The techniques used to evaluate the uncertainty factors are similar to i those used og CE plants employing Analog Reactor Protection Systems as  ;

described ',n References 2-1, 2-2 and 2-3. The only functional change is '

to ad4 the TheChSCUstochasticsimulationmethodologydescribedinReferences2-1 through 2-3 is used to determine the overall uncertainty factors.

2.3 TM/LP LSSS Stochastic Simulation For the TM/LP LSSS as described in Ref. 2-1, DNB overpower (Pfdn) is the dependent variable of interest. Core coolant inlet temperature, reactor coolant system pressure, core power, and peripheral axial shape index are monitored directly by the TM/LP trip system. Total integrated radial i peaking factor and RCS coolant flow rate are monitored by other systems l and must be included in the TM/LP LSSS evaluations.  ;

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Figure 2-1 is a flow chart representing the simulation sequence for the TM/LP LSSS. For each simulation trial, a value of overpower obtained using sampled valtes of uncertainties about nominal conditions is J calculated. This value is compared to the overpower calculated at nominal conditions by taking the ratio of the two values. This l simulation sequence is repeated over a large number sets of ncminal  !

cperating conditions ccvering the operations space for the plant. The l

. resulting distribution of the ratio of nominal overpower to overpowtr  !

incorporating uncertainties is used to detArmine the overall uncertainty j factor on the TM/LP LSSS. For the SAFDL, i This is done to be consistent with the requirements of the Section 4.S of the Standard Review Plan.

2.4 CNB LCO Stochastic Simulation  !

For the DNS LCO, DNS overpower (Pfdn) divided by the required margin (ROPM) is the dependent variable of interest. The core coolant inlet temperature, reactor coolant system pressure and flow rcte, peripheral axial shape index and integrated radial peaking factor are the indeperdent variables of interest. The CNBR SAFDL is included as a distributed input in en identical fashion to the TM/LP LSSS. Similar to 2-1 l l

i the approach taken for SCU (Ref. 2-1), the maximum R0PM as a function of shape index is used as input to generate the ONB LCO. This reduces the

' analytical evaluation of the dependent variable to consideration of Pfdn response to the uncertainties of the independent variables. CETOP-D is used to determine the functional relationship between Pfdn and the independent variables.

The probability distributions of uncertainties associated with the-independent variables have been discussed in Reference 2-3. <

Figure 2-2 is a flow chart representing the ex-core detector monitoring

. stochastic simulation of the DNB limits. This figure is similar to Figure 2-1.

2.5 References for Section 2:

2-1* "Statistical Combination of Uncertainties, Part 1" CEN-124(B)-P, December, 1979; 2-2* "Statistical Combination of Uncertainties Part 2. January, 1980" CEN-124 (B)-P, 2-3* "Statistical Combination of Uncertainties, Part 3" CEN-124 (B) P.

March 1980 2-4 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E),

"Regardin and SER)"g Unit

, June 24,1 1982.

Cycle 6 License Approval (Amendment 871 to OPR-53

• These References have been approved for use by the NRC in Appendix A of the Attachment to Reference 2-4.

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I 3.0 RESULTS AND CONU.USIONS 3.1 ' Results of Analysis Table 3-1 presents the TM/LP LSSS and DNS LCO statistically combined i uncertainties from application of the ESCU analytical methods described >

, in Section 2.

The individual uncertainties and their correspending values.which were combined are the same as presented in References 3-1 to 3-3 except for

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the addition of two uncertainty items to the MONBR p.d.f. (described in

Appendix A) which were added in the course of the NRC review.

The aggregate uncertainties are in units of and are applied as' described in Section 3.1.2 of Re"arence 3-1 for TM/LP LSSS and  ;

Section 3.1.1 of Reference 3-3 for DNS LCO.

3.2 Conclusion The Extended Statistical Combination cf uncertainties (ESCU) methods and typical results have been presented in this report. This methodology ,

will continue to provide at least a 95% probability at a 95% confidence level that the hottest fuel red will not experience Departure from Nucleats Boiling during normal operation or an Anticipated Operational Occurrence initiated from within the LCO limits. ,

3.3 References for Section 3:  :

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3-1* "Statistical Combination of Uncertainties part 1", CEN-124(B).P  !

Oecember, 1979.

3-2* "Statistical Cembination of Uncertainties Part 2", CEN-124(B)-P 4 January, 1980 ,

3-3* "Statistical Combination of Uncertainties Part 3", CEN-124(B).P March, 1980

! 3-4 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E). "Regarding  ;

d

. Unit 1 Cycle 6 License acproval (amendment 471 to OPR-53 and SER)",

June 24, 1982 t

• These References hava teen accroved for use by the NRC in Appendix A of ,

the Attach. tent to Reference 3-c.

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. TABLE 3-1 STATISTICALLY.COM81NED , )

UNCERTAINTIES FOR A  !

Appr.ox. values of Equiva, lent- j

-i TM/LP LSSS DN8 LCO EX-CORE i DNB LCO BAS $$

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i APPENDIX A A.0' MONBR pecbability Density Function A.1 Background A minimum DNBR probability density function (p.d.f.) was derived in Reference A-1 in order to arrive at the MONBR limit of 1.23. That limit accounted for uncertainties in system parameters and provided at least 95% probability and 95% confidence level that the hot fuel pin would not experience Departure from Nucleate Boiling (DNB).

This MONBR limit accounted for the following uncertainties:

a core inlet flow distribution ,

b engineering factor on enthalpy rise

• c systematic fuel red pitch d systematic fuel clad 0.0.

e engineering factor on heat flux  :

f CE-1 Critical Heat Flux (CHF) correlation t g fuel rod bow In course of their review (Reference A-2) of tha Statistical Combination of Uncertainties (SCU) methods described in Reference A-1, the NRC identified two additional uncertainties to be included in the system parameter SCU analyses:

h) CE-1 CHF correla' tion cross validation penalty (5% increase in CHF correlation standard deviation)

1) T-H code uncertainty penalty (5%. equal to two standard deviations).

I Subsequent to review and approval of the original Calvert Cliffs SC'J aaalysis, generic review of the CE-1 CHF correlation was completed by '

NRC. As a result of the generic review the tiRC approved the use of a i lower DNBR limit (1.15) for the C* 1 CHF correlation applied to CE's 14 x i 14 fuel rather than the 1.19 limit approved for CE's 16 x 16 fuel  !

(Reference A-3). Since the Calvert Cliffs SCU analysis included a CHF

, correlation uncertainty allowance corresponding to the 1.19 limit and Calvert Cliffs utilizes 14 x 14 fuel, the SCU MONBR limit was revised in the Calvert Cliffs I,' nit 2 Cycle 7 license submittal to credit the lewer CPF correlation uncertainty associated with the 1.15 CNBR limit. This change was approved by the NRC in Reference A-4

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A,2 NCNBA PDF j

The MONSR pdf is shewn in Figure I A-1. It inclu' des the follcwing system paramete'r uncertainties: I a core inlet flow distribution b engineering factor on enthalpy rise i c systematic fuel rod pitch ,

d) systeadtic fuel clad 0.0. l e) engineering factor on heat flux 1 f) CE-1 CHF correlatien l A-1

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. h) CE-1 CHF correlation cross validation penelty j

(5% increase in CHF correlation standard deviation) ,

1) T-H code uncertainty penalty (5%, equal to two standard 1 deviations). .

I

AsdiscussedinSectionA.1, uncertainties (a)through(g)were 1, considered in the original Calvert Cliffs system parameter SCU analysis (Reference A 1), while uncertainties (h) and (i) were added in the course-

, of the NRC review (Reference A-2) of the original SCU anc. lysis.. The CE-1  !

t CHF correlation uncertainty (g) used to derise tho MONOR p.d.f.

i- cor-esponds to the ONBR limit approved by 3 the NRC in Referenca A-4 4cr the CE-1 CHF correlation applied to CE's 14

, x 14 fuel and includes the CE-1 rHF correlation cross validation penalty j (h). ,

A rod bow pensity of 0.6% is included in the pdf. ,

H j This basedpenalty on CE's corresponds approved rod bow to an assembly methods (avera'ge Reference A-5). burnup of 45 GW/MTV

],

System parameter gicertainties and allowances acconmodated in the a pdf are presenta;d in Table A-1 alo,ng with the MON 8R  !

l mean and standard <1eviation of the p.d.f. ,

I j >

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Table A-1 Component System Parameter, Uncertainties and, Allowances Accomodated by pdf Standard Deviation Componeqt_, Uncertainty Mean at 95t confidence hot assembly inlet flow factor -

cnannel' ' inlet flow factor "l channel inlet flow factor channel- Inlet flow factor - - '

enthalpy rise engineering factor 1.00 0.010 systematic fuel red pitch (in) ' '

systematic fuel c1.sd 0.0, (in) , <

heat flux engineering factor 1.00 0.013 CE-1 CHF correlation [ )

Thermal Hydraulic code 1.00 0,025 uncertainty ov9rall PDNBR pdf i

annel r;urhers refer to Fig. A-2, originally presented as Figure 3-5 in

,, 8efertnce A-1. , -

Ststittias cased on CNBR limit and including 5% cross validatien tenalty on st edard 'devi'ation.

inclufu 0.37. red bow penalty.

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1 2 CHANNEL NUMBER IN FIRST STAGE MODEL 3 4 5 6 7 8 9 10 11 12 13 ,

14 15 16 17 18 19' 20 21 M 23 24 25 26 U 28 I

i I l  !

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- _. - h -. ._ __ . _. h _ - y __ - _.._L__._.. --

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CHANNEL NUMBERING SCHEME FOR STAGE 1 TORC ANALYSIS TO ESTABLISH RE'PONSE SURFACE STATE PARAMETERS A-5 G

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A.3 References for Aopendix A A-1. "Statistical Combination of Uncertainties Methodology Part 2:

Combination of System Parameter Uncertainties in Thermal l'argia Analyses for Calvert Cliffs Units 1 & 2", CEN-124 (B)-P, January, 1980.

A-2. Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (8G&E), "Regarding Unit 1 Cycle 6 License Approval (Ammendment #71 to OPR-53 and SER)",

June 24, 1982.

A-3. Letter, C. O. Thomas (NRC) to A. E. Sherer (CE), "Acceptance for o Referencing of Licensing Topical Report CENPD-207 (P/NP)", C-E Critical Heat Flux: Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids; Part 2-Non-uniform Axial Power Distribution",

November 2, 1984 A-4. Letter, D. H. Jaffee (NRC) to A. E. Lundvall (BG&E),

Subject:

Ammendment No. 9 to Facility Operating License No. DPR-69 for Calvert Cliffs Nuciear Power Plant, Unit No. 2, Oocket No. 50 - 318, November 21, 1985.

A-5. Fuel & Poison Rod Bowing", June, 1983, CENPD-225-P-A.

e d

I

• A-6 l

B.0 proposed Changes to the Technical Specification Bases B.1 Discussion )

As presented in the body.of this report, the Extended SCU analysis ha:

been performed so that a  !

assures with a 95% probability at a 95% confidence level l that the hottest fuel pin does not experience departure from nucleate j boiling during steady sti.te operations or anticipated operational ,

, occurrences. Therefore, . l

- l 1

based on the methods described in this report. To clarify this point in the Technical Specifications, the proposed Technical Specifications Bases will substitute the numerical value of the DNBR limit (1.21) with the term "the DNS SAFDL Consistent with the methods described in I CEN-348(B)-P". CEN-3a8(B)-P is the topical number for this report. j Table B-1 presents a list of the affected Technical Specifications Bases.

l e

S B-1

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_ TABLE B-1 Technical Specifications Bases Requiring Change Oue to incorporation of the ESCO Basis 2.1.1 (ReactorCore)-

Basis 2.2.1 (Reactor Coolant flow-low, axial flux offset, Thermal Margin / Low Pressure) 4 Basis 3/4.2.5 (Di') Parameters) 4 6

m B

1 e

i 4

B-2

, . . _ - . . ._ . _ . . _ . _ . _ , , . _ _ _ , . _ . . . . _ . . _ , _ . . _ _ _ . ,_.,___;_,,,.,_,_____