ML18138A048
| ML18138A048 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/21/1980 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| 134, NUDOCS 8002260464 | |
| Download: ML18138A048 (3) | |
Text
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VIRGINIA ELECTRIC A.ND POWER COMPANY RXCHMOND, VJ:RG'IN.XA 23261 February 21,. 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Albert Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Denton:
Serial No. 134 PO/FHT:scj Docket Nos. 50-280 50~281 License Nos. DPR-32 DPR-37 Asymmetric LOCA Loads on Reactor Pressure Vessels Surry Power Station Units 1 and 2 In response to Mr. Victor Stella's letter of January 25, 1978, Virginia Electric and Power Company became a participant in a Westinghouse Owners Group to resolve the subject issue.
The purpose of this letter is to review the status of the evaluation and to provide the results of the analyses completed to date.
The evaluation program was divided into three phases, A, B, and C.
Phase A included data acquisition from the Utilities, and a review o~ structural and hydraulic parameters for potential grouping of generically similar plants.
Phases Band C separated the evaluations for breaks postulated outside the reactor cavity and _inside the reactor cavity.
Phase B involved the actual structural assessments of plant groups and_ development of specific plant qualification programs as required for breaks outside the reactor cavity.
Phase C included evaluation of breaks inside the reactor cavity annulus and verification of the structural integrity of the reactor vessel *and supports, reactor internal structures, fuel, and ECCS piping attached to the reactor coolant system.
The integrity of the CRDM's and primary equipment supports which may be controlled by these vessel nozzle breaks is also considered in Phase C.
Concurrent with the Phase Band C work, mechanistic pipe break analyses were also undertaken to determine if large through-wall cracks in reactor coolant system piping would propagate to a large LOCA.
Results of this work have previously been submitted by Westinghouse letter NS-TMA-2200, dated February 6, 1980, as WCAP 9558, Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack."
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VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton 2
This report and the NSAC/EPRI Technical Memorandum submitted to the NRC on October 19, 1979, in a letter from John E. Ward (Chairman, AIF Committee on Reactor Licensing and Safety) to Harold R. Denton, have determined, by diverse and independent analyses and experimental results, that the probability of high energy line breaks in reactor piping systems, both austenitic and ferri-tic, is extremely small.
In addition the consequence of unanticipated, slow crack growth due* to fatigue, co.rrosion fatigue, or stress corrosion cracking is likely to be relatively slow leakage~
The analyses specifically determined that very large cracks ate required to initiate ductile fracture in nuclear piping under normal loadings; if ductile fracture does initiate due to a severe overload, unstable crack extension is unlikely to occur; and the open-ings of through wall cracks are small.
These results support the conclusion that a double-ended guillotine break in a reactor system pipe without any prior indication of substantial leakage is unrealistic and need not be considered as a basis for plant design or modifica-tion.
Nevertheless, Phase Band Phase C asymmetric loads analyses have been continued.
Results have been and will be submitted as described below.
Westinghouse Owners Group report "Phase B5: Subcompartment Asymmetric Pressure Loads", authored by D. S. Nixdorf was presented to your staff in February, 1979.
The remainder of the Phase B work, covering steam generator and reactor coolant pump integrity and supports evaluation, is reported in WCAP 9628, "Westinghouse Owners Group Phase B Asymmetric LOCA Loads Evaluation", which was.submitted by Westinghouse letter NS-TMA-2200 dated February 6, 1980.
Phase C results for verification of the structural integrity of the reactor vess~l supports and ECCS piping attached to the reactor coolant system are being submitted by Westinghouse letter NS-TMA-2206, dated February 14, 1980.
In accordance with the agreement reached with the NRC staff in November, 1979, results of evaluations of reactor internal structures, fuel and control rod drive mechanisms will be provided by July, 1980.
In addition, the Westinghouse Owners Group has analyzed two representative plants and presented the results to the NRC in a meeting on February 21, 1979.
The representative plants analyzed used nominal existing plant configurations.
including break limiting devices in the reactor cavity shield wall.
The above analysis results have been compiled because of the s~aff's express desire to gain a better understanding of the asymmetric loads issue.
We co.ntinue to believe the additional Mechanistic Fracture Evaluation work which the Owners Group undertook provides sufficient justification to eliminate the double-ended guilotine break as a basis for plant design.
We urge that review of WCAP 9558 continue and that its conclusions be adopted as resolution of this issue.
We do not feel that backfitting the Surry plant will provide substantial additional protection of the public health and safety.
To the contrary, modification will impose additional radiation exposure to those installing the modifications~
VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold Denton 3
The NRC has stated that asymmetric LOCA loads should be combined with seismic loads.
This position is not justified as demonstrated by the Mechanistic Fracture Evaluation.
We agree with the NRC Staff assessment "that the probability of a pipe break resulting in substantial transient loads on the vessel support system or other structures is acceptably small because: (1) the break of primary concern must be very large, (2) it must occur at a specific location, (3) the break must occur essentially instantaneously, and (4) the welds are currently subject to inservice inspection by volumetric and surface tech-niques in accordance with ASME Code Section XI".
Therefore operation of Surry Units 1 and 2 is justified while this matter is being resolved.
cc:
Mr. James P. O'Reilly Very truly yours, C. M. Stallings Vice President-Power Supply and Production Operations