ML17055D655
| ML17055D655 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/08/1988 |
| From: | Haughey M Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-66539, NUDOCS 8803140230 | |
| Download: ML17055D655 (80) | |
Text
March 8, 1988 Docket No.
50-410 LICENSEE:
FACILITY:
FACILITY:
DISRUBTION--~
tOO~T NRCPDR Niagara Mohawk Power Corporation (NMPC) Local PDR PDI-1 Rdg.
Nine Mile Point, Unit 2 RCapra CVogan MEETING
SUMMARY
FOR JANUARY 20, 1988 MEETING WITH MHaughey NIAGARA MOHAWK POWER CORPORATION AND WESTINGHOUSE HBClayton ELECTRIC CORPORATION ON WESTINGHOUSE FUEL EJordan JPartlow TCollins DFieno JJohnson OGC-WF ACRS(10)
Because a recent agreement between Westinghouse and General Electric may limit the use of Westinghouse barrier-design fuel for boiling water reactors (BWRs),
NMPC is considering its options with respect to what fuel will be purchased for use in upcoming core reloads for Nine Mile Point, Unit 2.
On January 20, 1988, the NRC staff met with representatives of NMPC and Westinghouse Electric Corporation (W) concerning the use of Westinghouse fuel at Nine Mile Point, Unit 2.
Because of lsmited staff resources, the NRC stated that it would be unable to complete a review of more than one fuel design on a schedule that would support the reload schedules for Nine Mile Point, Unit 2.
Furthermore, if a decision on the type of fuel was not made shortly, the NRC might not be able to complete a
review of a single design within the required schedule.
The NRC had halted review of any topical report on W BWR fuel that was affected by potential changes in fuel design and was awaiting a determination from NMPC concerning which fuel design would be used before resuming the reviews.
In addition, the NRC staff expressed that if other BWR plants on a tight schedule chose to use a fuel design different than NMPC, then the NRC would evaluate which design review would accommodate the most plants and give priority to that design.
During the meeting W noted that it was negotiating with three other utilities for alter nate designs.
Copies of slides presented during the meeting from both NMPC and W are included as Enclosure l.
A list of meeting attendees is included as Enclosure 2.
Sincerely,
Enclosures:
As stated cc:
See next page PDI-1,p CVogan~
3/q /88 P
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M g ey:dig 3/
/88 Mary F. Haughey, Project Manager Project Directorate I-l Division of Reactor Projects, I/II q.M PDI-1 RCapra 3/8 /88 SSPgi40230 85000410 ssp~oS PDR ADQCK 0 PDR)
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'1 Il I II fA Iff r
If If II I
Mr. C.
V. Mangan Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit 2 CC:
Mr. Troy B. Conner, Jr.,
Esq.
Conner A Wetterhahn Suite 1050 1747 Pennsylvania
- Avenue, N.W.
Washington, D.C.
20006 Mr. Richard Goldsmith Syracuse University College of Law E. I. White Hall Campus
- Syracuse, New York 12223 Mr. Don Hill Niagara Mohawk Power Corporation Suite 550 4520 East West Highway
- Bethesda, MD 20814 Resident Inspector Nine Mile Point Nuclear Power Station P. 0.
Box 99
- Lycoming, New York 13093 Mr. Gary D. Wilson, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West
- Syracuse, New York 13202 Mr. Peter E. Francisco, Licensing Niaqara Mohawk Power Corporation 301 Plainfield Road
- Syracuse, New York 13212 Regional Administrator, Region !
U.S. Nuclear Regulatory Comnission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Paul D.
Eddy New York State Public Serice Commission Nine Mile Point Nuclear Station-Unit 2 P.O.
Box 63
- Lycoming, New York 13093 Mr. Richard M. Kessel Chair and Executive Director State Consumer Protection Board 99 Washington Avenue
- Albany, New York 12210 Mr. Richard Abbott, Unit 2 Station Superintendent Nine Mile Point Nuclear Station Niagara Mohawk Power Corporation P. 0.
Box 32
- Lycoming, NY 13093 Mr. Thomas Perkins, General Supt.
Nine Mile Point Nuclear Station Niagara Mohawk Power Corporation P. 0.
Box 32
- Lycoming, NY 13093 Ms.
Donna Ross New York State Energy Office 2 Empire State Plaza 16th Floor
- Albany, New York 12223
A
WESTINGHOUSE/NIAGARA MOHAWK POWER CORPORATION/NRC MEETING JANUARY 20, 1988 MEETING AGENDA 0
INTRODUCTION 0
WESTINGHOUSE BWR FUEL GENERIC LICENSING PROGRAM 0
SUMMARY
/ CONCLUSIONS
l
('
WESTINGHOUSE / NIAGARA MOHAWK POWER CORPORATION /
NRC MEETING JANUARY 20, 1988 INTRODUCTION 0
BACKGROUND
- WESTINGHOUSE/GE BARRIER CLAD SETTLEMENT
- WESTINGHOUSE BWR FUEL OBLIGATIONS AND ALTERNATIVES 0
PURPOSE OF MEETING
- REVIEW WESTINGHOUSE BWR FUEL GENERIC LICENSING NEEDED IN SUPPORT OF CURRENT WESTINGHOUSE BWR FUEL, SUPPLY OBLIGATIONS
- REVIEW IMPACT OF FUTURE WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES ON CURRENT BWR FUEL GENERIC LICENSING TOPICALS 0
MEETING OBJECTIVES
- AGREE THAT THE CURRENT WESTINGHOUSE BWR FUEL GENERIC TOPICAL REPORTS ARE APPLICABLE TO THE CURRENT WESTINGHOUSE BWR FUEL OBLIGATIONS, AS WELL AS, FUTURE WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES (WITH MINOR MODIFICATIONS/SUPPLEMENTS)
- REAFFIRM NEED FOR CONTINUED NRC REVIEW OF CURRENT WESTINGHOUSE BWR FUEL GENERIC TOPICAL REPORTS REQUIRED TO SUPPORT THE CURRENT WESTINGHOUSE BWR FUEL OBLIGATIONS
C 0 g
WESTINGHOUSE / NIAGARA MOHAWK POWER CORPORATION /
NRC MEETING JANUARY 20, 1988 WESTINGHOUSE BWR FUEL GENERIC LICENSING PROGRAM 0
GENERIC LICENSING PROGRAM IN SUPPORT OF CURRENT WESTINGHOUSE BWR FUEL SUPPLY OBLIGATIONS 0
IMPACT OF FUTURE WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES ON CURRENT BWR FUEL GENERIC LICENSING PROGRAM
GENERIC LICENSING PROGRAM IN SUPPORT OF CURRENT WESTINGHOUSE BWR FUEL SUPPLY OBLIGATIONS 0
WESTINGHOUSE BWR FUEL GENERIC TOPICAL REPORT LICENSING PROGRAM 0
NINE MILE POINT UNIT 2 QUAD+ LTA LICENSING PROGRAM
L
Summary of Westinghouse BWR Topical Reports To ical Submittal Date To NRC Requested NRC SER Date Nuclear Design Codes (WCAP-10106)
{PHOENIX/POLCA)
Thermal-Hydraulic Code (MCAP-10107)
{CONDOR)
Nuclear Qualification (WCAP-10841)
{PHOENIX/POLCA) 8x8 CPR Correlation (MCAP-10972)
{MQB1)
Channel Stability Code (WCAP"11146)
(MAZDA-NF)
Non-LOCA Transient Code Qualification (WCAP-11236)
(BISON)
LOCA Code Qualification (MCAP-11284)
(GOBLIN/CHACHA/DRAGON)
QUAD+ CPR Correlation (WCAP-11287)
(MB1)
LOCA Sensitivity Studies (MCAP-11427)
BWR Fuel Rod Design (MCAP-11503)
(PAD BMR)
W BWR Fuel Reference Safety Report TMCAP"11500)
Rod Drop Methodology Core Stability Code (NUFREQ-NP)
Fuel Assembly Seismic/LOCA Evaluation 6-82 6"82 6"85 6-86 8"86 9-86 6-87 6-87
'8"87 12-87 12-87 12-~.
Completed Completed 2"88 3"88 4-88 8-88 9-88 3-88 10-88 3-89 3-89 3-89 3-89
0
WESTINGHOUSE / NIAGARA MOHAWK POWER CORPORATION /
NRC MEETING JANUARY 20, 1988
SUMMARY
/ CONCLUSIONS 0
CURRENT WESTINGHOUSE BWR FUEL GENERIC TOPICAL REPORTS ARE APPLICABLE (WITH 85~) MODIFICATIONS/SUPPLEMENTS TO 4 OUT OF 14 TOPICALS)
TO FUTURE WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES 0
CONTINUED REVIEW OF CURRENT WESTINGHOUSE BWR FUEL GENERIC TOPICAL REPORTS IS NECESSARY TO SUPPORT CURRENT WESTINGHOUSE BWR FUEL SUPPLY OBLIGATIONS
l
IMPACT OF FUTURE WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES ON CURRENT WESTINGHOUSE BWR FUEL GENERIC LICENSING PROGRAM 0
DESCRIPTION OF WESTINGHOUSE BWR FUEL DESIGN ALTERNATIVES
- QUAD+ 4X4 W/0 LINER
- QUAD+ 5XS W/0 LINER 0
IMPACT OF QUAD+ 4X4 W/0 LINER ON CURRENT TOPICAL REPORTS 0
IMPACT OF QUAD+ 5X5 W/0 LINER ON CURRENT TOPICAL REPORTS
IMPACT OF QUAD+ 5X5 0 CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 10106 - WESTINGHOUSE NUCLEAR DESIGN CODES DESCRIPTION
= (PHOENIX/POLCA)
PURPOSE
- DESCRIPTION OF WESTINGHOUSE BWR NUCLEAR METHODS 5X5 IMPACT-
IMPACT OF QUAD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 10107 - WESTINGHOUSE THERMAL-HYDRAULICCODE DESCRIPTION (CONDOR)
PURPOSE - DESCRIPTION OF WESTINGHOUSE BWR THERMAL-HYDRAULICMETHODS 5X5 IMPACT-
IMPACT OF QUAD+.5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP10841 -
WESTINGHOUSE NUCLEAR DESIGN CODES QUALIFICATION (PHOENIX/POLCA)
PURPOSE QUALIFICATION/BENCHMARKOF WESTINGHOUSE BWR NUCLEAR DESIGN METHODS 5X5 IMPACT-
/IN'~Pi~ ~~ jQ e~Qgg~~+u/Q(Wo~,Fe G 7o)
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IMPACT OF AD+ 5X5 ON C
RRENT WESTI GH USE BWR TOPICAL REPORTS WCAP 10972 - WESTINGHOUSE CPR CORRELATION FOR 8X8 FUEL (WQB1)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF A WESTINGHOUSE CORRELATION FOR PREDICTING CPR MARGINS IN BWRS FOR GE 7X7, GE 8X8, GE 8X8R, GE P8X8R, AND ASEA ATOM 8X8 FUEL BUNDLE DESIGNS 5X5 IMPACT - A/On/8
IMPACT F
AD+ 5X 0
CURRENT WESTINGHOUSE BWR TOPICAL REPORT WCAP 11146 -
BWR FUEL BUNDLE CHANNEL STABILITY ANALYSIS CODE (MAZDA-NF)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF WESTINGHOUSE BWR FUEL BUNDLE CHANNEL STABILITY ANALYSIS METHODS 5X5 IMPACT - h'W'~.
I
IMPACT OF QUAD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11236 - NON-LOCA TRANSIENT ANALYSIS CODE (BISON)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF WESTINGHOUSE BWR NON-LOCA TRANSIENT ANALYSIS METHODS 5X5 IMPACT -
dA/6
IMPACT OF QUAD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11284 -
WESTINGHOUSE BWR LOCA ANALYSIS CODES (GOBLIN/CHACHA/DRAGON)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF WESTINGHOUSE BWR LOCA ANALYSIS METHODS 5X5 IMPACT-
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IMPACT OF AD+ 5X5 ON CURRENT WESTINGHOUSE BWR T PICAL REPORTS WCAP 11287 - WESTINGHOUSE QUAD+ FUEL BUNDLE CPR CORRELATION (WB1)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF A WESTINGHOUSE CORRELATION FOR PREDICTING CPR MARGINS IN BWR CORES FOR QUAD+ 4X4 FUEL BUNDLES 5X5 IMPACT-
lI 0
IMPACT OF QUAD+ 5X5 0 CURRENT WESTI GHOUSE BWR TOPICAL REPORTS WCAP 11427 - WESTINGHOUSE BWR LOCA ANALYSIS METHODS APPENDIX K SENSITIVITY STUDIES PURPOSE
- DESCRIPTION OF WESTINGHOUSE METHODOLOGY FOR PERFORMING LOCA SENSITIVITY STUDIES AND DEFINITION OF THE APPENDIX K EVALUATION METHODOLOGY. DEFINITION OF AN APPENDIX K EVALUATION MODEL FOR A TYPICAL'WR/5 PLANT IS ALSO PRESENTED ALONG WITH THE WESTINGHOUSE METHODOLOGY FOR ASSESSING THE LOCA IMPACT ON MIXED CORES 5X5 IMPACT-
V
IMPACT OF AD+ 5X5 ON CURRE T WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11503 - WESTINGHOUSE BWR FUEL ROD PERFORMANCE MODELS (PADBWR)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF WESTINGHOUSE BWR FUEL ROD PERFORMANCE MODELS 5X5 IMPACT -
dW :
4 o
IMPACT OF Q AD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11500 - WESTINGHOUSE BWR FUEL REFERENCE SAFETY REPORT PURPOSE
- DESCRIPTION OF THE WESTINGHOUSE QUAD+ 4X4 MECHANICAL,
- NUCLEAR, AND THERMAL-HYDRAULICDESIGN, WESTINGHOUSE METHODOLOGY FOR DESIGNING BWR FUEL BUNDLES (MECHANICAL, NUCLEAR, THERMAL-HYDRAULIC), AND THE WESTINGHOUSE METHODOLOGY FOR PERFORMING RELOAD SAFETY ANALYSIS ON BWR CORES 5X5 IMPACT-
I i
~
IMPACT OF UAD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11685 - WESTINGHOUSE BWR CONTROL ROD DROP ACCIDENT ANALYSIS PURPOSE
- DESCRIPTION OF WESTINGHOUSE BWR CONTROL ROD DROP ACCIDENT ANALYSIS METHODOLOGY 5X5 IMPACT - A+sN-
T I
~
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IMPACT OF Q AD+ 5X5 0 C RRENT WESTINGHOU E
BWR TOPICAL REPORTS WCAP 11684 - WESTINGHOUSE BWR CORE STABILITY ANALYSIS CODE (NUFREQ-NPW)
PURPOSE
- DESCRIPTION AND QUALIFICATION/BENCHMARKINGOF WESTINGHOUSE BWR CORE STABILITY ANALYSIS METHODS 5X5 IMPACT -
8'PW
I
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IMPACT OF QUAD+ 5X5 ON CURRENT WESTINGHOUSE BWR TOPICAL REPORTS WCAP 11686 - SEISMIC/LOCA EVALUATION OF WESTINGHOUSE QUAD+ 4X4 FUEL PURPOSE
- DEFINITION OF WESTINGHOUSE METHODOLOGY FOR EVALUATION OF BWR FUEL ASSEMBLIES SUBJECTED TO SEISMIC/LOCA LOADS (DESIGN BASES, STRESS CRITERIA, ACCIDENT DEFINITION),
DESCRIPTION OF ANALYTICALMETHODS, QUALIFICATION/BENCHMARKINGOF FUEL BUNDLE MECHANICAL
- MODELS, AND EVALUATION OF THE WESTINGHOUSE QUAD+ 4X4 BWR FUEL DESIGN 5X5 IMPACT-cy (s~/i~~P z~~od/j
a a
S MMARY OF WESTINGHOU E
D UMENTATION IMPACTED BY QUAD+ 5X5 0
CRITICAL POWER CORRELATION 0
LOCA APPENDIX K SENSITIVITY STUDIES IOo 0
FUEL ASSEMBLY SEISMIC/LOCA EVALUATION 0
BWR FUEL REFERENCE SAFETY REPORT o W'4~ C~ ~~~
c - rv~M ~>~< ~~
CA~.~-A~
r C
44 VER D+ 4X4 R
R ELATI WB-PI AL 0
ASPECTS WHICH WILL NOT CHANGE RELATIVE.-TO 5X5
- METHODOLOGY"FOR EVALUATION OF CRITICAL POWER TEST DATA
- METHODOLOGY FOR DEVELOPMENT OF THE CPR CORRELATION
- METHODOLOGY FOR EVALUATION OF THE CPR CORRELATION 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5
- TEST RESULTS
- DEVELOPMENT OF THE CPR CORRELATION
- EVALUATION OF THE CPR CORRELATION
P 0
4X4 VERS 5X5 LOCA APPENDIX K SE SITIVITY STUDIES 0
ASPECTS WHICH WILL NOT REQUIRE MODIFICATION FOR 5X5 ANALYTICALMETHODS DEFINITION OF APPENDIX K EVALUATION METHODOLOGY REFERENCE LOCA DESCRIPTION PLANT PARAMETERS SENSITIVITIES CODE NUMERICAL STUDIES METHODOLOGY FOR TRANSITION CORE EVALUATION 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5 MODEL CHANGES FOR CHF AND TWO-PHASE PRESSURE DROP SENSITIVITIES TO SPRAY HEAT TRANSFER,
- CCFL, NODING, POWER SHAPE DISTRIBUTION, LIMITING BREAK SPECTRUM, MIXED CORES
4 VE EI MI L
EVAL ATI O
ASPECTS WHICH WILL NOT REQUIRE MODIFICATION FOR 5X5
- DESIGN BASES
- STRESS CRITERIA
- COMPUTER CODES
- ANALYTICALMETHODS
- ACCCIDENT DEFINITION AND ASSVMPTIONS CORE PLATE MOTION LOCA DEFINITION 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5
- INPUT TO FINITE ELEMENT MODEL SOME BENCHMARK TEST RESULTS
- MECHANICAL RESPONSE OF ASSEMBLY TO ACCIDENT CONDITIONS FUEL ROD AND CHANNEL STRESS ASSEMBLY LIFT
4 l
4X4 VERSUS 5X5 IMPACT ON RSR CHAPTER 2 - MECHANICAL DESIGN ASPECTS WHICH WILL NOT REQUIRE MODIFICATION FOR 5X5
- DESIGN BASES
- DESIGN CRITERIA
- ANALYTICALMETHODS COMPUTER CODES
- STEADY-STATE AND TRANSIENT LOADING ASSUMPTIONS
- FAILURE MECHANISMS
- QA, TESTING AND POST-IRRADIATION INSPECTION PROGRAMS
- CORROSION/HYDRIDING EVALUATION
- COMPATIBILITY EVALUATION 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5
- FUEL ROD PERFORMANCE RESULTS
- MECHANICAL RESPONSE OF ASSEMBLY TO STATIC, FATIGUE AND SHIPPING LOADS
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~ ~
$ ~
4X4 VER X
IMP CHAPTER 3 - NUCLEAR DESIGN 0
ASPECTS WHICH WILL NOT REQUIRE MODIFICATION FOR 5X5
- DESIGN BASES
- COMPUTER CODES
- ANALYTICALMETHODS POWER DISTRIBUTION CALCULATIONS, LOADING PATTERN DETERMINATION, REFERENCE CORE APPROACH, SLCS AND SDM METHODS
- UNCERTAINTIES
- METHODS FOR ESTABLISHING INPUT TO SAFETY'NALYSES
- NUCLEAR BEHAVIOR OF THE CORE 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5
- SAMPLE BUNDLE CHARACTERISTICS
~ C
4X4 VER S 5X5 IMPA T 0 RSR CHAPTER 4 - THERMAL AND HYDRAULIC DESIGN 0
ASPECTS WHICH WILL NOT REQUIRE MODIFICATION fOR 5X5
- DESIGN BASES
- COMPUTER CODES
- METHODOLOGY CRITICAL POWER RATIO, SAFETY LIMIT 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5
- PRESSURE DROP DISTRIBUTION WITHIN BUNDLE
- SPECIAL ANALYSIS SUPPORTING SAFETY LIMIT
4X4 VERSUS 5X5 IMPACT ON RSR CHAPTER 5 - RELOAD FUEL SAFETY ANALYSIS 0
ASPECTS WHICH WILL NOT REQUIRE MODIFICATION FOR 5X5 SAFETY EVALUATION APPROACH COMPUTER CODES EVENT FREQUENCY DEFINITIONS EVENT CATEGORIZATION SELECTION OF BOUNDING TRANSIENT EVENTS METHODOLOGY FOR BOUNDING TRANSIENTS EVALUATION OF BOUNDING TRANSIENT EVENTS DESCRIPTION OF SPECIAL EVENTS 0 OVERPRESSURE 0
SHUTDOWN MARGIN 0 STABILITY 0 LOCA 0 CRDA METHODOLOGY FOR SPECIAL EVENTS DESCRIPTION OF UTILITY OPTIONS METHODOLOGY FOR UTILITY OPTIONS 0
ASPECTS WHICH WILL REQUIRE MODIFICATION FOR 5X5 EXPANDED DISCUSSION OF STABILITY
- STABILITY RESULTS
- SENSITIVITY STUDY RESULTS
- OPERATING LIMIT MCPR UNCERTAINTY ANALYSIS
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1.2 CONCLUSION
S This report concludes that the design bases defined for the evaluation of the Nestinghouse QUAD+ BMR fuel assembly are adequate to ensure compliance with all regulatory criteria, including the General Design Criteria of 10CFR50 Appendix A, as they are applicable to fuel systems and the effect of reload fuel on reactor systems.
Included among these design bases are fuel design criteria which represent Specified Acceptable Fuel Design Limits (SAFDLs) which meet the guidelines of Section 4.2 of the Standard Review Plan (NUREG-0800) for satisfying the objectives of GDC10.
In addition, sufficient information has been provided to assure that the requirements and guidelines of NUREG-0800, which are applicable to reload fuel and core designs, can be sa ti sf ied for reloads incorporating QUAD+ fuel.
The report also concludes that the methodology described and demonstrated is
- adequate, for design and licensing applications, to ensure that the design bases are met when QUAD+ BHR fuel assemblies are inserted into any U. S.
commercial BMR/2 through BWR/6.
A.
The results of the Mechanical Design evaluation performed on the QUAD+
BNR fuel assembly confirmed that:
o The design bases identified are sufficient to assure that all requirements and guidelines identified in section 4.2 of NUREG-0800 wi 11 be satisfied.
o The methodology described for evaluating the thermal fuel performance, mechanical, and materials behavior of the QUAD+ BHR fuel assembly relative to the design bases is acceptable for licensing and design purposes.
~,C
o The QUAD+ BRR fuel assembly evaluations provide an illustration of the methodology to be utilized for each application of QUAD+ BMR
. fuel assemblies.
These evaluations also demonstrate that the QUAD+
assembly meets the fuel performance, mechanical, and materials design bases under normal operational, anticipated operational occurrences, and accident conditions.
B.
The results of the Nuclear Design evaluations performed for BNR cores containing 8x8 and QUAD+ fuel assemblies confirmed that:
o The nuclear design bases and functional requirements are sufficient to assure that all requirements and guidelines identified in Section 4.3 of NUREG-0800 will be satisfied.
Typical core arrangements have been analyzed to demonstrate that the design bases can be satisfied when QUAD+ fuel is inserted in a reload core.
The nuclear design characteristics of cores containing QUAD+ fuel (reactivity coefficients, control rod worths and power distributions) are sufficiently similar to currently operating BMR cores that acceptable margins to safety limits can be demonstrated.
o The nuclear design computer codes and models described in this and other referenced topical reports can be applied in the manner described to determine the neutronic behavior of BMR fuel.
o The QUAD+ BNR fuel assembly design does not adversely affect core operation and monitoring functions.
C.
The results of the Thermal-Hydraulic Design evaluation of the QUAD+ BHR fuel assembly and cores containing QUAD+ assemblies confirmed that:
o The design bases identified are sufficient to assure that all requirements and guidelines of Section 4.4 of NUREG-0800 will be satisfied.
RllS:d b1NOT 1-7
~ E
o The gUAO+ BMR fuel assembly is hydraulically compatible with non-Restinghouse 8x8 assemblies and can co-reside with them in transition cores.
o Sufficient thermal-hydraulic testing has been performed to develop correlations for accurately predicting gUAD+ hydraulic and thermal-hydraulic performance.
o The thermal performance design basis and methodology described for establishing the Safety Limit Hinimum Critical Power Ratio is applicable to cores containing Westinghouse SQUAD+ fuel assemblies and non-Westinghouse fuel assemblies.
o For a typical QUAD+ BMR reload core, a safety limit minimum Critical Power Ratio (CPR) value of 1.07 calculated from the MB-1 CPR correlation ensures that more than 99.9/ of the fuel rods in the reactor are expected to avoid dryout.
o The thermal-hydraulic design computer codes and models described in this and other referenced topical reports can be applied in the manner described to determine the thermal-hydraulic behavior of B'HR fuel.
D.
It is concluded in the Reload Safety Analysis section of this report that:
o The methodology presented and illustrated by example calculations is applicable'o the safety analysis of BWR plant designs BMR/2 through BAR/6 and will satisfy the requirements and guidelines of Sections 5.2.2 and 15 of NUREG-0800 applicable to reload fuel.
o Although as many as thirty different transient and special events must be considered for evaluation with new reload fuel the detailed analysis of a subset of these will constitute an acceptable safety analysis.
3b73f:5 V0407 1-8
r~
o Hased on the information presented in this report it has been determined that a reload incorporating the OUAD+ fuel assembly and applying the Westinghouse safety analysis methodology for reload cores does not represent any unreviewed safety questions as defined by 10CFR50.59.
- Further, any inssuance of a license amendment for the installation and use of (jUAO+ fuel does not involve a signifi-cant hazards consideration.
1-9
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NMPC/NRC MEETING'NMP2 RELOAD FUEL JANUARY 20 1888 PURPOSE.'
TO INFORN NRC OF THE NMP2 RELOAD FUEL SITUATION ALTERNATIVES BEING CONSIDERED PRESENT OVERALL SCHEDULE HEEDS
r
CHRONOLOGY OF EVENTS 9/10/87,'GENERAL ELECTRIC INITIATES BARRIER PATENT LAWSUIT AGAINST WESTINGHOUSE 12/9/87.'ESTINGHOUSE AHD GENERAL ELECTRIC SETTLE PATENT DISPUTE 12/9/87',
WESTINGHOUSE INFORMS NMPC OF SETTLEMENT AGREEMENT 12/11/87,'HMPC INFORMS NRC OF PATENT SETTLEMEHT 1/4/88'NMPC/WESTINGHOUSE UPDATE NRC AND SCHEDULE MEETING 1/20/88.'HNPC/WESTINGHOUSE/NRC MEETING
J pC IP 0
NNPC ACTION CONTINUE TO NEET PRESENT CONTRACTUAL OBLIGATIONS EVALUATING THE SITUATION
I E MILE P INT UNIT 2 QUAD+ LTA LI E SING D CUMENTATI DOCUMENT SUBMIT TO NRC SER SCHEDULE WCAP 10507 - QUAD+ DEMONSTRATION ASSEMBLY REPORT COMPLETE RECEIVED WCAP 11159 - SUPPLEMENTAL QUAD+ DEMONSTRATION ASSY COMPLETE REPORT FOR FITZPATRICK RECEIVED WCAP 10841 - WESTINGHOUSE NUCLEAR CODE QUALIFICATION 6/85 2/88 WCAP 10972 - 8X8 CPR CORRELATION WCAP 11236 - NON-LOCA TRANSIENT CODE QUALIFICATION WCAP 11287 - QUAD+ 4X4 CPR CORRELATION WCAP 11503 -
PADBWR FUEL ROD PERFORMANCE MODELS 1/86 8/86 9/86 6/87 3/88 9/88 3/88 10/88 NINE MILE POINT UNIT 2 RELOAD LICENSING SUBMITTAL (INCLUDING QUAD+ LTA'S) 5/89 8/89
J i<
Attendance List January 20, 1988 Enclosure 2
Nary Haughey Tim Collins Daniel Fieno Don Hill G.
K. Rhode K. W. Korcz Bob Kraus Wayne Hodges Robert A. Capra Paul E. Netusil Allan McFarlane J.
Pat Duruet Bin Harris Ron Carr Doug Bevard ORGANIZATION NRC - Licensing Project Manager NRC - Reactor System Branch NRC - Reactor System Branch NNPC - Licensing NMPC - Consultant NMPC - Licensing NNPC - Fuels NRC - SRXR NRC - PDI-1 NNPC - Lead Nuclear Fuel Engineer Westinghouse Westinghouse BWR Programs Westinghouse -
BWR Engineer.
Westinghouse BWR Programs'estinghouse
- Nuclear Safety
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