ML20127C183
| ML20127C183 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/12/1984 |
| From: | Asselstine J NRC COMMISSION (OCM) |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML20126G811 | List: |
| References | |
| FOIA-84-885 NUDOCS 8410230414 | |
| Download: ML20127C183 (12) | |
Text
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.# UNITED STATES NUCLEAR REGULATORY COMMISSION e,h f
ng jj waswiwcrow.c.c.nosss , V**ow/ *** October 12, 1984 CFFICE OF THE COMMISSIONER , MEMORANDUM TO: Files FROM:
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d - James K. Asselstine/ i
SUBJECT:
MEETING WITH REPRESENTATIVES OF SOUTHERN CALIFORNIA EDISON - RESTART OF SAN ONOFRE 1 On October 10, 1984 I met with Don Craven, Dave Piggott, Jim Altman and Ken Baskin, representatives of So~uthern California Ed,ison, to discuss legal issues associated with restart of San Onofre 1. The Company representatives stated that they had not read the Comission's 1982 order as amending the license. The issue of license amendments was not raised until June of this year by the staff, and it surprised the company. The company asserts that the Comission has the flexibility to treat the order as not being an amendment. They see no legal barrier to this interpretation and significant policy implications if we do not interpret this order correctly. Further, the equities call for a flexible approach. This was a voluntary upgrade, and the company has relied on consistent staff interpretation that no hearing was required. I said that I was troubled by the staff's interpretation of the '82 order because it might limit our discretion in future enforcement matters and make F voluntary compliance unlikely in the future. However, I asked if this case wasn't closer to Sho11y than to the license suspension cases like Bellotti. The company representatives tried to distinguish the Sho11y case. They said that in Sholly the Comission's order granted authority not consistent with the suspended license. In ti.is case, however, the upgrade of the plant is not barred by the San Onofre license. No change in tech specs is required and there is no unreviewed safety question so conduct of the upgrade is consistent with the license. ! My legal assistant then asked if the .5g design basis earthquake requirement is a part of the license or incorporated into it somehow. Mr. Baskin and Mr. l Piggott said they thought it might be, but would have to check. I explained i that I was concerned that if the .5g requirement was a license requirement, and the staff did not think that the plant met that requirement then allowing the plant to operate at some lesser or indeteminate level might constitute a license amendment. Mr. Baskin explained that there is substantial evidence I which indicates that the plant is designed to withstand .5g. Further, the i cuestions in the minds of the staff are not based on substantial evidence and are based on questionable techniques. Licensee intends to provide additional information to the staff, but practically they cannot go back and recreate the design. The question will be what standard of evidence is necessary for a finding that there is reasonable assurance that the plant does meet .5g. Ifd[ [dY}h M w w.
m av - The Southern California representatives then asked if they had allayed my concerns. I said that as far as allowing restart before all the upgrade to .67 was concerned I did not have' major legal problems. I wanted to look at the staff's technical analysis, however. My major concern was whether the plant w met .5g and whether .5g is a part of the license. Again, I said I .ould have to see the staff's technical analysis. I also told them that I had a particular technical concern with not upgrading the saltwater intake. Mr. Baskin said that wasn't a problem because there were several other sources of water available - a water tank, a reservoir and fire trucks. The Southern California representatives said they would keep in touch with my legal assistant to see if there were any additional questions they could answer or information they could provide.
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4 9 UEU U E 3CI I 215N SAN ONOFRE 1 SEISMIC REEVALUATION FOR RESTART The " return to service" (RTS) scope proposed by the Southern California Edison Company (SCE) for San Onofre Unit 1 includes all safety-related structures, the reactcr coolant pressure bcundary, the main steam and main feedwater system boundary, capability to provide reactor coolant system makeup and capability to remove decay heat through the steam generators (to a hot standby condition) using the auxiliary feedwater system. The scope includes control power (nitrogen or el'ectrical) for the required equipment. The spent fuel pool will remain intact, and provides a source of borated water for reactivity control, as well as an additional limited source of cooling water. In an evaluation dated February 8, 1984, the staff concluded that the scope of seismic upcrades, which SCE proposed to complete, would be sufficient to ensure safe plant shutdown to a hot standby condition following a design-basis earthquake (0.679) modified Housner ground response spectra. The seismic capability for the balance of plant systems and components, including cold shutdcwn capability, would be resolved in the SEP integrated assessment. The staff is presently preparing a St.fety Evaluation Report for restart which acdresses implementation cf the RTS plan based on information developed throuch meetings, site inspections and analyses audits. The staff is issuing a recuest to SCE to decurren.t the findings necessary to cc:rplete the 57fety Ev-luation Report. The staf f's basis fer ccnchdtnc that phrt cperation v.culd not pcre an undue ritt +r
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y* , lJ i .L , ,. . l* - 2-d i including isolation boundaries necessary to ensure sufficient integrity i i to remove decay heat, have been upgraded to 0.67 9 ; (3) the systens which 3 - would remove decay heat through either the steam generators, with active 4 f redundancy E, have been upgraded to 0.67 9 ; and (4) the auxiliary feedwater i i storage tank and spent fuel pool have been upgraded to 0.67 9, and would 1 provide a cooling water supply and' supplementary reactivity control (withcut 1 ) impairing safe spent fuel storage), respectively, to maintain the reactor in a j i hot standby condition for several days or more. 5. sed on these provisions, J, ! the staff concludes that it is unlikely that a significant seismic event could i 1 4 occur during the time that it will take to complete the resolution of the I l seismic design issue, which would cause an accident or transient requiring the i function of the systems for which seismic analyses and upgrading are not yet cceplete.' Therefore, the plant would be able to achieve hot standby for a , j si;nificant seismic event, at least up to 0.67 9. This cooling water supply i j wculd allow sufficient time to assess plant damage from such an event and develop a I centingency plans to provide sufficient cooling viater to achieve cold shutdcwn, i Fcr syste:rs and components not within the RTS sccpe, which principally ccnsist 1
- of the accident mitigatien systems, some upgrading has been completed. During i !
i 15EE and 1983 reedifications were made ,to parts of these systens. Experience i j , , has been developed as part of USI A-46, " Seismic Qualifict. tion," and generic { tiging desigr studies which woulo suggest that, even thcugh the analyses nd i J
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tih_% - 00T 12133.; p . i upgrading has not been completed, these systems would likely survive a significant seismic event and, with some minimal repairs, could perform their j intended functions. However, an explicit correlation between this experience
- and the San Onofre Unit I systems, sufficient to quantify their seismic
- capability, would be almost as extensive and time-consuming as the analyses i j necessary to resolve this issue. 'Therefore, despite the improvements made and the related experience, the staff cannot unquestionably. conclude at this time, that these systems would or would not properly function following a 0.679 earthquake.
j Based on a finding that either (1) the Confirmatory Order established a licensed i seismic design basis of 0.679 , or (2) a sufficient basis has been established i to invalidate the original seismic design basis of 0.5g for these systems, exemptions or a license amendment may be necessary to authorize plant bperation without a complete, established seismic capability for the accident mitigation , ) systems. The licensee will formally be requested to identify and justify any I necessary exemptions; they preliminarily have indicated that they do not agree with either of the findings by which exemptions would be necessary. Nevertheless, the staff has screened the regulations and plant license in an effort to identify the exemptions and license arendments that would be necessary, given that either 1 . . j finding is made, with particular empha' sis on the GDC. The staff has not identified I- eny licer.se arnendments associated with restart, other than that associated with j the August 1982 Confirr.atory Order for seismic modifications. 4 l 1 a 4 i
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- uni;'. c OCT 12 ;gge i ! It is important to recognize that several issues have been identified which could similarly be related to regulatory requirements. Because these issues have not been the subject of orders or license conditions nor has a basis been i developed to invalidate the original licensing basis, the staff has not ! cersidered that' exemptions for these issues would be required (e.g., conteinment isolation configuration). This approach is consistent with that'taken on other SEF plants. Criterion 2 - Design Basis for Frotection Against Natural Phenomena In the discussion below, GDC 2, applies for all systems which are not within the 0.TS scope. i The arplicable requirements of GDC 2 are " structures, systems and components important'to safety shall be designed to withstand the effects of natural 3 chencmena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and I seiches without loss of capability to perform their safety functions. The l cerign bases for these structures, systems and components shall reflect: (1) accropriate censideration of the most severe of the natural phenomena that i { have.been historically reported for the site and surrounding area, with l
- I sufficient margin' for the limited accuracy, quantity, and period of time in
] 1 which the historical data have been accumulated, (2) appropriate enrbinatior.s l Of the effects of r.0rr.al and accident ccr.ditions with the effects of the r a'.et i phencr2la and (3) the irportance of the safety functions to be
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Lq~hi t_ 00T 12 595 j, f Discussion The basis for both 0.679 and 0.5g have been adequately documented; the latter in relation to the original license. As previously described, the staff believes j that an adequate technical basis exists to conclude the plant can safely shutdown for such events, until the design of the remaining systems and
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components important to safety have been established and/or upgraded to 0.679 . 1 I Systems Used for Reachino Cold Shutdown Residual Heat Removal System s Co .;:or.ent Cooling Water System
; Circulating Water (Salt Water Cooling) System Critericn 34 - Residual Heat Removal ;
i "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits 4 and the design conditions of the reactor coolant pressure boundary are not exceeded." t . Criterion 44 - Cooling Weter "A 5es*.en to transfer heat from structures, syr.tcr.s cr.d creponents important tr tr'ety to an ultipate her.t sink shall be previded. The syftet safety
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Discussion The return to service plan includes cooling capability to reach and maintain a hot standby cor.dition by adding auxiliary feedwater to the steam generators and exhausting the steam to the atnosphere or, if necessary, " feeding and bleeding" the prinary system for longer-term-cooling with external water sources. This will permit the plant.to'be held in a stable condition while repairs are made to the above systens or until other sources of water could be obtained for longer-term cooling and eventual cold shutdown. Accident fliticatine Systems Safety injection system, including recirculation mode Containment spray, including chemical addition Cc.ntainment Isolation Criterion 4 - Environmental and Missile Design Bases
" Systems, structures and compenents important to safety shall be designed to accc=odate the effects of conditions associated with... postulated accidents, including less-of-coolant accidents. These structures, systems and comperents shall be app ~roprietely protected against dynanic effects...that may result fron events and conditions cutside the r.uclear power unit."
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%.I2 s i 1 Criterion 35 - Emergency Core Cooling "A systen to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following 4 any loss of reactor coolant..." Note that, in this case, 10 CFR 50.46) does i not apply because the fuel is clad in stainless steel (Interim Acceptance j
Criteriaapply). 1 Criterion 38 - Containment Heat Removal 3
"A systen to remove heat from the reactor containment shall be provided. The ; system safety function shall be to reduce rapidly...the containment pressure
{ and terperature following any loss-of-coolant accident." l Criterion 41 - Containment Atmosphere Cleanup and 10 CFR 50.44 s
" Systems to control fission products...and other substances [ including hydrogen) which may be released into containment shall be provided as necessary to reduce...the concentration and cuality of fission products released to the environment following postulated accidents."
4 Criterion 50 - Containment Design Basis )
"The reac'.or containment structure and the containment heat removal system shall i be designed such that the centainment structure can accommodate without exceedir.g the desigr. leakage rate and with sufficient margin, the calculated trett.u s and ter.iperature ccnditions resulting frer Any loss-of-crolan; eccident..."
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-c-Criterion 54 - Piping Systems Penetratir.g Containment
" Piping systems penetrating containmer.t shall be provided with...isolaticn capabilities having... reliability ard performance capabilities which reflect the incortance to safety of isolating these piping systems..."
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Discussion GDC 4 relates seismic events to loss-of-coolant accidents. The other criteria relate to systems required to mitigate loss-of-coolant accidents. For the return to service plan the objective was to ensure integrity of components whose failure could lead to a LOCA, to preciude the need for LOCA-related systems. The react 5r coolant pressure boundary and the boundary of main steam and main
'eed.ater piping have been upgraded to withstand the 0.679 earthcuake. This scope also encompasses some portions of the containment isolation functions fcr lir,es directly connected to the reactor coolant system.
Since ciping whose failure could lead tc a loss-of-coolant accident has beer, u;; reded to withstand an 0.679 earthquake, the likelihood of a seismically-ir.d;.:ed LOCA requiring use cf accident-r'tigating systens is srall. 5:c-- sni Pcci Coolir;
.' <'inr 61 - Fuel Stm age and Handi' c and P,adicettivity Certr-1
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m a g r---+m.~m-*+%>-- -* m T' "'M " -~ "-- -** ~ I ' I ** %l UNITED STATES *A. NUCLEAR REGULATORY COMMISSION j jNf;pf '^j WASHINGTON, D. C. 20555 d' March 5, 1984 ....+ MEMORANDUM FOR: Frank J. !!iraglia, Assistant Director for Safety Assessment, DL FROM: Dennis M. Crutchfield, Chief Operating Reactors Branch iP5, DL
SUBJECT:
TELEPHONE CONVERSATION WITH CALIFORNIA PUBLIC UTILITIES COPMISSION - SAN ONOFRE UNIT 1. On February 23, 1984 L' alt Paulson, Eileen McKenna and I had a telephone conversation with nembers of the staff of the California Public Utilities Connission (PUC) concerning the hearings scheduled to begin on February 27, 1984 on whether or not San Onofre Unit 1 should be renoved from the rate base. The focus of the conversation was the written testimony filed by David Fogarty on behalf of Southern California Edisen Company before the PUC. We indicated that Fogarty's testimony appeared to be an accurate representation of the facts. The PUC staff then asked several questions relating to the following:
- 1) Schedule for completion of seismic modifications and staff review for restart.
- 2) What actions nust be completed by the NRC staff and by the licensee for restart
- 3) Extent of NRC staff review of the analysis results, modifications and implementation.
- 4) The interrelationship of the integrated living schedule with the seismic restart program.
- 5) The irpact of plant nodifications of the different requirements for short and lo,ng-tern operation as well as of staff exceptions to the restart criteria.
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- 6) Any possible new reauirements that could be imposed for restart.
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- 7) Costs associated with modification.
We noted that the schedule will be basically determined by the licensee's ability to complete the analysis and nodifications. Fogarty's testinony states that SCE's pro,iection for return to operation is by Decenber 31, 1984 The na.ior-nnn-seismic issues needed for restart are technical specifications and the steam generator inspection progran. N $ i S['. ] 2 ' 'f; ,
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Frank J. Miraglia - 2- March 5, 1984 The PUC staff seemed particularly concerned with quality control of the inplementation of the modifications. We noted the role of the licensee's quality assurance program and of the regional inspection efforts. In the area of costs the PUC staff was referred to previous licensee submittals as well as to the filed testimony. MIN $ - Dennis M. Crutchfield, hief Operating Reactors Br ch #5 Division of Licensing cc: .E; McKenna L. Chandler D. Eisenhut . W. Paulson 4 1}}