NUREG-0609, Forwards Request for Addl Info on Generic Ltr 84-04,Safety Evaluation of Westinghouse Topical Repts Dealing W/ Eliminating of Postulated Pipe Breaks in PWR Primary Main Loops. Response Requested by 840731
| ML17215A433 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/31/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | Williams J FLORIDA POWER & LIGHT CO. |
| References | |
| RTR-NUREG-0609 GL-84-04, NUDOCS 8406140271 | |
| Download: ML17215A433 (9) | |
Text
Docket No. 50-335 DISTRIBUTION:
~ocket File NRC PDR L PDR Mr. J.
W. Williams, Jr.
ORB//3 Rdg Vice President DEisenhut Nuclear Energy Department OELD Florida Power 8 Light Company EJordan P. 0.
Box 14000 JNGrace Juno Beach, Florida 33408 ACRS-10 Gray File
Dear Mr. Williams:
PMKreutzer
SUBJECT:
ASYMMETRIC LOCA LOAOS OPERATING REACTOR LICENSING ACTIONS-
SUMMARY
OF CONCERNS OF INDEPENDENT PLANT SAFETY EVALUATIONS Re:
St. Lucie Plant, Unit No.
1 Generic Letter 84-04, dated February 1,
1984 subject "Safety Evaluation of Westinghouse Topical Reports Dealing With Eliminating of Postulated Pipe Breaks in PWR Primary Main Loops" indicated an acceptable technical basis has been provided for the 16 Westinghouse Owners Group plants so that the asymmetric blowdown loads resulting from double ended pipe breaks in the main coolant loop piping need not be considered as a design basis provided certain specified conditions are met.
- However, as you are one of several independently represented nuclear plant facilities currently under review for their ability to withstand asymmetric loadings from postulated loss-of-coolant accident (LOCA), we are transmitting this request for additional information in order to complete our Safety Evaluations which are scheduled for completion in the fourth quarter of fiscal 84.
Under review are the plant-specific LOCA analysis submitted by the independent utilities for the following plants:
Salem Unit 1, Trojan, Beaver Valley Unit 1, Prairie Island Unit 1 and 2,
- Kewaunee, Maine Yankee and St. Lucie Unit 1.
The asymmetric LOCA load submittals reviewed to date are referenced at the end of the enclosed summary.
All submittals were evaluated with the guidelines set forth by NUREG-0609 and the summary lists the major areas of concern that have not met these established guidelines.
Detailed questions and requests have previously been submitted in the form of initial reviews and additional requests for information.
It is our understanding that a
substantial portion of the information which would resolve these concerns is available with the NSSS vendors.
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Nr. Williams Please respond to the concerns identified in the enclosure in time for the staff to complete the Safety Evaluation as scheduled above, i.e.,
by July 31, 1984.
The reporting and/or recordkeeping requirements of this letter affect fewer than ten respondents; therefore, ONB clearance is not required under P.L.96-511.
Enclosure:
As stated Qriginal Signcd bY j R~
James R. Niller, Chief Operating Reactors Branch ¹3 Division of Licensing cc w/enclosure:
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Florida Power
& Light Company CC:
Harold F. Reis, Esquire Newman
& Holtzinger 1025 Connecticut
- Avenue, NW Washington, DC 20036 Norman A. Coll, Esquire McCarthy, Steel, Hector'nd Davis 14th Floor, First National Bank Building Miami, Florida 33131 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Mr. Weldon 8. Lewis County Administrator St. Lucie County 2300 Virginia Avenue, Room 104 Fort Pierce, Florida 33450 U.S. Environmental Protection Agency Region IV Office ATTN:
Regional Radiation Representative 345 Courtland Street, HE Atlanta, Georgia "30308 Mr. Charles B. Brinkman Manager - Washington Nuclear Operations C-E Power Systems Combustion Engineerinq, Inc.
7910 Wondmont Avenue
- Bethesda, Maryland 20014 Regional Administrator Nuclear Regulatory Commission Region II Office of Executive Director for Operations 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Mr. Jack Schreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 Resident Inspector c/o U.S.
NRC Senior Resident Inspector 7585 S.
Hwy AIA Jensen
- Beach, Florida 33457 State Planning
& Development Clearinghouse Office of Planning
& Budget Executive Office of the Governor The Capitol Building Tallahassee, Florida 32301
REQUEST FOR ADDITIONAL INFORMATION 6.
ST.
LUCIE 1
'OCKET NO. 50-335 A. Cavit Pressurization Anal sis 1.
A sensitivity study is required to demonstrate conservatism of the nodalization used in the RELAP-3 analysis of the reactor vessel cavity.
2.
A discussion is required about the vent areas that become available as a result of fluid flow or pressurization effects after initiation of the accident.
The areas must be justified analytically or experimentally..
3.
The calculated forces and moments acting on the reactor vessel and shield wall, resulting from the reactor vessel cavity
- analysis, are required in quantitative form.
4.
A complete description is required of the derivation and quantitative data for the cavity pressure loads used to evaluate the St. Lucie internals and fuel.
Application to the St. Lucie plant must be demonstrated.
5.
The CONTEMPT input listing and description used in the steam generator subcompartment analysis is required.
6.
A description of the method used for determing the blowdown data utilized in the steam generator subcompartment analysis is required.
B. Thermal H draulics Anal sis 1.
The postulated pipe break locations, break size, and break opening times are required for the analyses performed using RELAP-4.
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2.
Additional nodalization and modeling details are required for the RELAP-4 models, identify~ng the various components and dimensional details.
3.
A list of RELAP-4 thermal hydraulic input parameters is required for each break analyzed.
N 4.
The resulting absolute and differential pressure transients across the core barrel are required for each break analyzed.
5.
References and documentation of the CEFLASH-4B analyses performed are required to show their application to the St.
Lucie plant for the evaluation of the fuel and internals.
C. Structural Anal si s 1.
The following information is required to clarify the primary coolant 'system model:
- a. detailed schematics which exhibit all of the structural elements and nodalization b.
A description of how system support stiffnesses were determined and implemented in the model 2.
A more detailed and substantial justification is required to demonstrate the adequacy of the four steam generator sliding base holddown bolts that exceed the design loads.
3, A more substantial justification is required to demonstrate the adequacy of the overstressed reactor coolant pump discharge nozzle, especially considering the 7.78 ft break loadings.
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4.
A complete LOCA analysis adhering to the criteria outlined in NUREG-0609 is required for tive primary shield wall and steam generator subcompartment wal'ls.
5.
A description is required of the differences and similarities
'etween the internals and fuel components of the generic analysis and the St. Lucie 1 plant.
"Reactor Coolant System Asymmetric LOCA Loads Evaluation",
Revision 1,
Enclosure to letter L-80-263, Florida Power and Light Company, August 8, 1980.
St.
Lucie Unit No.
1 Final Safety Analysis Report, Amendment No.
36, Docket No. 50-335, December ZO, 1974.