ML19296B762
| ML19296B762 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/13/1980 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8002210480 | |
| Download: ML19296B762 (4) | |
Text
,
O Wisconsin Electnc ma couesur 231 W. MICHIGAN P.O. BOX 2046. MILWAUMEE Wa 53201 February 13, 1980 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention: Mr. A.
Schwencer, Chief Operating Reactors Branch 1 Gentlemen:
DOCKET NOS. 50-266 AND 50-301 ASYMMETRIC LOCA LOADS EVALUATION POINT BEACH NUCEEAR PLANT, UNITS 1 AND 2 Victor Stello's letter of January 25, 1978, requested that we evaluate asymmetric loads developed during a Loss of Coolant Accident (LOCA) and provide a schedule for completion of the evaluation.
A Westinghouse Owners' Group was formed to complete an evaluation of the type requested and developed a two year schedule for the program.
The evaluation program was divided into three phases; A, B, and C.
Phase A included data acquisition from the utilities, and review of structural and hydraulic parameters for potential grouping among generally similar plants.
Phases B and C separated the evaluations for breaks postulated outside and inside the reactor cavity respectively.
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 area.
Phase C included evaluation of breaks inside the reactor cavity annulus and verification of the sturctural integrity of the reactor vessel and supports, reactor internal structures, fuel, and ECCS piping attached to the reactor coolant system.
Concurrent with the Phase B and C work, mechanistic pipe break analyses were undertaken to determine if large through-wall cracks in reactor coolant system piping would propagate, allowing a double-ended rupture of a main coolant pipe.
Results of this work have previously been submitted by Westinghouse for O
. 8002p1 04-949
Mr. Harold R.
Denton February 13, 1980 the Owners' Group in the form of WCAP-9558
(. Proprietary) and WCAP-9570 (Non-Proprietary), " Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack", letter NS-TMA-2200, T. M. Anderson to D.
G. Eisenhut, dated February 6, 1980.
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 ferritic, is extremely small.
The analyses specifically determined that very large cracks are 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 openings of through-wall cracks are small.
Therefore, the consequence of unanticipated, slow crack growth due to fatigue, corrosion fatigue, or stress corrosion cracking is likely to be a relatively small amount of leakage.
These results support the conclusion that a double-ended guillotine break in a reactor coolant system pipe without any prior indica-tion of substantial leakage is unrealistic and need not be considered as a basis for plant design or modification.
Nevertheless, Phase B and Phase C asymmetric load analyses have been continued.
Results have been and will be submitted as described below.
Westinghouse Owners' Group report,
" Phase BS: Subcompartment Asymmetric Pressure Loads", authored by D. S. Nixdorf, was submitted for Staff review in February, 1979.
The remainder of the Phase B work covering steam generator and reactor coolant pump integrity and support evaluation is reported in WCAP-9628 (Proprietary) and WCAP-9662 (Non-Proprietary),
" Westinghouse Owners' Group Asymmetric LOCA Loads Evaluation Phase B", which was submitted by Westinghouse on February 6, 1980.
Phase C results verifying the structural integrity of the reactor vessel and supports and ECCS piping attached to the reactor coolant system have been submitted by Westinghouse for the Owners' Group (letter NS-TMA-2206, T. M. Anderson to D.
G. Eisenhut, February 14, 1980).
In accordance with an agreement reached with the NRC Staff in November, 1979, final results of the Phase C evaluation will be provided by July, 1980.
In addition, the Westinghouse Owners' Group also analyzed two
" representative" plants and presented the results to the NRC at a meeting on February 21, 1979.
The " representative" plant analyses used nominal existing plant configurations.
The specific plant analyses for Point Beach Nuclear Plant, Units 1 and 2, assumed a modified plant configuration, including for example, break limiting devices in the reactor cavity shield wall.
Mr. Harold R.
Denton February 13, 1980 These asymmetric load analyses have been continued solely because of the Staff's expressed desire to gain a better understanding of the asymmetric loads issue.
We continue to believe that the additional mechanistic break work which the Westinghouse Owners' Group undertook presents sufficient justi-fication to preclude double-ended guillotine breaks as a basis for plant design or accido t analysis.
We urge that review of the mechanistic break topicol report, WCAP-9570 (WCAP-95 58,
Proprietary), be continued and that its conclusions be adopted as the basis for the resolution of this issue.
It is not clear that backfitting the modifications assumed in the asymmetric load analysee would ".
. provide substantial, additional protection which is required for the public health and safety.
, the criteria for backfitting as stated in 10 CFR 50.109.
To the contrary, a large cost burden would be placed on the utilities and their customers to implement these modifications.
In addition, radiation exposures would be incurred by plant operating and construction personnel during the backfit work without any demonstrated corresponding benefit.
Should the Staff require that the plant be modified to meet the analysis assumptions, despite the evidence presented in the mechanistic break report, the modifications could be installed during the second refueling outage after an agreement is reached on a full resolution of the asymmetric loads issue.
Two refueling cycles are required to install the modifications because detailed measurements must be made of the reactor cavity shield wall / reactor coolant pipe annulus prior to fabrication of break limiting devicer.
The NRC Staff has also stated that asymmetric LOCA loads must be combined with seismic loads.
We feel that this position is not justified as demonstrated by the mechanistic break work.
Furthermore, we know of no acceptable analytic technique for combining our existing elastic seismic loads with the inelastic asymmetric LOCA loads which were developed for some break locations.
If the Staff insists upon load combination, work to develop an acceptable technique must be initiated and the design of modifications would necessarily be postponed until such a technique is developed.
We agree with, and our analysis supports, the NRC Staff
(.r Stello's letter to All PWR Licensees dated assessment M
January 25, 1978, Enclosure 1), "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 techniques
Mr. Harold R.
Denton February 13, 1980 in accordance with ASME Code Section XI".
Therefore, continued reactor operation is certainly justified while this matter is being resolved.
Very truly yours, C. W.
Fay, Director Nuclear Power Department