ML20245C077
| ML20245C077 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/08/1989 |
| From: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Papay L SOUTHERN CALIFORNIA EDISON CO. |
| Shared Package | |
| ML13323A481 | List: |
| References | |
| NUDOCS 8906260054 | |
| Download: ML20245C077 (4) | |
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' ATTACHMENT:TO QUESTION 2:
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FEB8 1999 Docket No. 50-206 Southern California Edison. company. P. 0.-Box 800 2244 Walnut Grove Avenue Roseme'ad, California 92770 ; Attention: Dr L. T. Papay Senior Vice President Gentlemen:- Subjer.: Technical' Issues Impacting San Onofre Unit 1 Restart Reference is made to our Confirmatory Action Letter dated January 31, 1989, which documented our understanding that SCE will resolve questions concerning - T Unit I thermal shield integrity and other pending technicts issues before restart from the current refueling outage. The confirmatory action letter stated that these other technical issues would be the subject of separate correspondence. The purpose of this letter is to identify the additional i technical issues which the NRC has detemined must be resolved before-restart. As discussed by phone with you and members of your staff on February 7,1989, the NRC considers the following technical issues to require correction or other appropriate resolution before Unit I restart (i.e., entry into Mode 2):
- 1. The single failure problem which could potentially prevent automatic isolation of non-safety related loads on 480-volt swing bus No. 3.
- 2. Improper safety classification / environmental qualification of steam generator level instruments (inconsistent with post-TMI design commitment).
- 3. Environmental qualification of the south charging pump motor _ and other safety related motors which may have been previously rewound.
- 4. The presence of nonconforming piping (Schedule 120 vice 160) in the residual heat removal (RHR) system.
- 5. Inappropriatedesignofcomponentcoolingwater(CCW)valveswhich control CCW flow to the RHR heat exchangers.
- 6. Potential overloading of No. I and No. 2 480-volt switchgear in the event of a safety injection without a loss of offsite power.
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r 2 7. The adequacy of SCE's actions regarding repair and evaluation of leaking and/or improperly rolled steam generator tube sleeves. In addition to the above specific technical issues, a significant NRC concern emphasired during our February 7 discussion involves the number of single failure, environmental qualification (EQ), and design deficiencies which continue to be identified at Unit 1. We understand from our discussion that SCE shares this concern. As a result of the original single failure issue (reported in October 1987), escalated EQ enforcement issue (March 1988), and the findings of the NRC's safety system functional inspection (June 1988), SCE committed San Onofre. to reorgantre and reassess the perfonnance of engineering work at has identified several additional significant deficiencies.Since These include this re most of the specific technical issues listed above. The NRC is concerned that the total scope of these various deficiencies'is not being properly put into perspective by SCE. You are therefore requested to address this concern before restart. In particular. SCE is requested to provide to this office in writing, no later than seven days before scheduled-Unit I restart,including: deficiencies, an evaluation of recently identified Unit I design A listing of all significant Unit 1 single failure, EQ, and design deficiencies identified since enhanced design review processes were implemented (August 1988). An evaluation of the adequacy of your actions to correct the underlying root causes of the various deficiencies, in the context of Unit I readiness for restart. An evaluation of the effectiveness of previous reviews in view of the recently identified deficiencies, and actions which may be required to revalidate acceptable plant design. A definition of the scope of evaluations remaining to be completed which are likely to identify additional significant design deficiencies. Address the potential impact of these remaining evaluations on Unit I readiness to restart. We recognire that a number of desigti changes are also in progress to correct the single failure concerns reported in October 1987 and to address other NRC commitments connected with restart from the Cycle 10 refueling outage. We request that you also provide to us, before Unit I enters Mode 2, written certification that all coinmitments made to the NRC for actions to be completed during the Cycle 10 refueling outage have in fact been completed. _ _ _ _ _ - _ _ _ - - - - - - - - - ~ ~
t i m, - ';of :.,. ., , E-If other issues are identified which require resolution before Unit I restart, these will be discussed with you and will be the subject of further
. correspondence, if appropriate. Should you have any questions ~or* coments concerning the above, we will be pleased to discuss them with you .
1 cerely .
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J. B. Martin Regional Administrator cc: D. J. Fogarty, Executive Vice President C. P. K. B. McCarthy, Jr.,President Baskin, Vice Vice President - Site Manager (San Clemente) . H.E. Morgan,StationManager(SanClemente) State of California
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i 1 QUESTION 3 When exactly is NRC proposing to allow Edison to restart the San Onofre 17 ANSWER 1l On May 15, 1989, the NRC staff issued License Amendment No. 127, which estab-lished conditions for operation of San Onofre Unit I until the thermal shield is repaired. These conditions-include the establishment of a thermal shield
- monitoring _ program, a mid-cycle inspection, and development of a plan for repairing the thermal' shield during the next refueling outage. On May 16,
, 1989, the Director of the Office of Nuclear Reactor Regulation issued a letter to SCE indicating that the additional matters identified in the NRC's February 8 letter had been resolved to the staff's satisfaction and that the staff concurred in the licensee's plan to restart Unit 1. These two documents, which are attached, authorized the licensee to restart San Onofre Unit 1, which occurred on May 21, 1989.
.+ js* "%[4 L., " 9*TATEs ATTACHMENT T0 QUESTION 3 1 y ;Ty (;.gg NUCLEAP. ' .-RY COMMISSION q l wr.3..s :on.o.c.20sss
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, l l l Mr. Kenneth P. Baskin Vice President Southern California Edison Company. ]!
2244 Walnut Grove Avenue Post Offico Box 800 -l Rosemead, Ct.lifornia 91770
Dear Mr. Baskin:
SUBJECT:
ISSUANCE OF AMENDMENT NO. 127 TO PROVISIONAL OPERATING LICENS SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 (TAC NO. 71853) The Commission has issued the enclosed Amendment No.127 to Provisional Operating License No. DPR-13 for San Onofre Nuclear Generating Station, Unit No. 1. The anendment consists of changes to the operating license in response to your and Mayapplication dated February 17,1989, as supplemented March 21 and 23, 3 and 8, 1989. The amendment provides for a reactor vessel thermal shield monitoring program and mid-cycle inspection until the thermal shield fasteners are repaired during the fuel cycle XI refueling and 10-year ASME Inservice Inspection. SCE is requesteo to oevelop and submit for staff approval a conceptual design and plan for the repair of the thermal shield. Specific subjects tha? should be addressed are plans for the existing bolts that are not broken, the design of new flexures and modification to the limiter keys. Please provide this { information within 90 days. ; When the thermal shield and core barrel are removed, an nitrasonic test of the support block ledge should be perfomed from the inside af the core barrel. An , J 1 l . _- -- - -- - -- - )
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Mr. Kenneth P. Baskin May 15, 1989 ultrasonic examination should also be performed on a representative sample of_ the core barrel-to-lower support plate weld. A visual examination should be performed of the entire weld from both the inside and outside surface. Copies of our related Safety Evaluation and Notice of Issuance are also enclosed. Sincerely. 0 lA L f/: /r Charles M. Tramell Senior Project Manager Project Directorate V . Division of Reactor Projects III, IV, Y and Special Projects
Enclosures:
- 1. Amendment No. 127 to License No. DPR-13
- 2. Safety Evaluation 3 Notice of Issuance cc w/enclos n es:
See next page l l l
3 Mr. Keimeth P. Baskin l: Southern California Edison Cor.pany San Onofre Nuclear Generating Station, Unit No. 1 l Cc Mr. Kenreth P. Baskin Vice President Southern California Edison Company p 2244 Walnut Grove Avenue ! Post Office Box 800 ) Rosesead, California 91770 l- David R. Pigot?. Orrick,~ Herrington & Sutcliffe
- l. 600 Montgomery Street San Francisco, California 94111 l
.Mr. Robert G. Lacy.
Manager, Nuclear San Diego Gas & Electric Company - P. 0. Box 1831 San Diego,~ Califorth 92112 l- Resident Inspector / San Onofre NPS U.S. NRC P. O. Box 4329 San Clemente, California 92672 Mayor 1 City of San Clemente San Llemente, California 92672 Chairman Board of Supervisors County of San Diego 1600 Pacific highway Room 335 San Diego, California 92101 Regional Administrator, Region V U.S. Nuclear Regulatory Comission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Mr. Paul Szalinski, Chief Radiological Henith Branch State Department of Health Services 714 P Street, Office Bldg. #8 Sacramento, California 95814
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SOUTHERN CALIFORNIA EDIS0h COMPANY SAN DIEGO GAS AND. ELECTRIC COMPANY DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.127 License No. DPR-13
- 1. The Nuclear Regulatory Comission (the Comission) has founo that:
A. The application for amenoment by Southern California Edison Company and San Diego Gas and Electric Company (the licensee) dated February 17, 1989, as supplemented !! arch 21 and 23, and May 3 ano 8, 1989 complies with the standards and requirements of the
' Atomic Energy Act of 1954, as amenced (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be cenoucted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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Accordingly, the license is amended by the addition of new paragraph-3.M as indicated in the attachment to this license amendment. 3. This license amendment is effective as of the date of its issuance and must be fully implemented as cescribed in the Attachment. FOR THE NUCLEAR REGULATORY COMv!SSION-
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George . Knighto Director Project Directorate V . Division of Reactor Project.s III,
r, Y and Special Projects Office of Nuclear Reactor Regulation-
Attachment:
License Condition 3.M Date of Issuance: May 15, 1989 I j
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l 3.M cvela Y Thermal shield Monitorine PrssrAm The neutron monitor the noise condition of the
/ loose-parts detection system shall'. be used to throughout Cycle.X or until repair. reactor vessel . thermal shield Periodic. monitoring.of both'
' neutron noise and loose-parts vibrations confirms that no long tern unacceptable trend of degradation is occurring. The details of this program are described below.
(1) The unit will be shut down no later than June 30, inspect the condition of the thermal shield. 1990 to (2)- During the first 7 days of 1 854 power, interim acceptance critaria' for neutron noise / loose-parts monitoring will be developed. , These final acceptance interim criteria criteria vill be utilized until the is developed. Final acceptance criteria for neutron noise / loose-parts monitoring will be established by. performing baseline
. evaluations for 45 . calendar days at 1. 85% power following return to service for Cycle X operation. The base line data will be established by recording a minimum of 16 segments of j data information, each of 20 minute duration.at 1 85% power.
Adjustments to the acceptance criteria will be made for cycle burnup and boron concentration changes throughout the cycle. (3) The neutron noise / loose-parts monitoring system shall be i OPERABLE in MODE 1 with: a) At least two horizontal loose parts detectors monitored for at least five (5) minutes 2 times per day; and, b) at least three (3) neutron noise inputs monitored for at least twenty (20) minutes once a week, and be analyzed for cross 1 i coherence. power spectral density', including phase and i (4) The data provided by the loose-parts / neutron noise monitor l shall be analyzed established criteria. once per week and compared with the criteria If the data exceeds the acceptance a) Within 1 day tne NRC will be informed of the exceedance. b) Within 14 days the conditions will be evaluated and a report provided to the NRC documenting future plans and actions. Amendment No. 127 i l
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Thedemonstrated be plant will be' failed. shutdown should the remaining flexure (5) Each channel of the loose-part detection system shall be demonstrated OPERABLE in MODE 1 by performance of as a) CHANNEL CHECK at least once per 24 hours b) CHANNEL TEST at least- once per 31 days The surveillance requirements for neutron noise monitor are thw Power Range Neutron Flux.cr/.'ered by the Appendix A Technical S
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(6) With the neutron inoperable noise for more than/ loose-parts 7 days detection instrumentation , Report to the . Commission pursu,antlicensee shall submit a Special Specification 6.9.2 within the. nextto Appendix A Technical 3 days outlining the cause of system the malfunction to operable status. and the plans for restoring the (7) In the case indicated of asensors, on site seismic event of 0.25g or greater as initiated. Before operationsa controlled shut down shall be are resumed, it will be de=enstrated to the seismicthat no thermal shield damage has occurred due event. (8) The provisions of Appendix A Technical specification 3.0.4 are not applicable to this license condition. Amendment No.127
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SAFETY EVALUATION BY TH,E OFFICE OF WUCLEAR REACTOR REGULATION RELATED TO. AMENDMENT.HO.127TO PROVISIONAL OPERATING LICEN SOUTHERN _ CALIFORNIA EDISOLCOMPANY SAN.DIEGO GAS.AND ELECTRIC COMPANY SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NO. 50-206 1.0 _ INTR DUCTION By letter cated February 17, 1989, as supplemented March 21 and 23, and May 3 and 6, 1989, Southern California Edison Company (SCE or the licensee) requested a change to Provisional Operating License No. DPR-13 for opa: ration of San Onufre Nuclear Generating Station, Unit No.1, located in San Diego County, California. 2.0 DISCUSSION In response to an alert from the reactor vendor that reactor vessel thermal shield fssteners at another facility had been found degradeo, SCE inspected the thermal shield for SONGS-1 on January 3-4, 1989. The thermal shield at San Onofre Unit I surrounds the reactor core barrel. It is 21" thick and about 10 feet in height, and weighs 48,000 lbs. It is supported at the bottom by six support blocks and thirty bolts which attach it to the core barrel. Support at the top is provioed by six flexures and 4 limiter keys. Five of the six flexures have been known to have been broken since 1978. j The licensee described the inspection results and presented video tape recordings at a January 27, 1989 meeting. During the last refueling outage visual examinations were performed with the internals installed using a high resolution underwater television camera system. Although the core was loaded during the inspection, selected fuel assemblies were shuffled to provide access, The licensee confirmed that five of six thermal shield flexure fixtures are broken as detected in a previous inspection. Three out of thirty thermal shield support block bolts are broken. bolts in support blocks at the O' and 240' locations.TheseNo areother the 7/8-inch signiti- top cant degradation was observec visually. Ultrasonic testing was not performed. O c6 99 l
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1 The licensee has evaluated operation with the thermal shield in the ' conoition worst observed expected, worstand considered credible, three and worst cases which it characterizes as: conceivable. The worst expected case involves the degradation of bolts at the third i support block, and the sixth flexure remaining intact. The worst credible case assumes that all: support blocks degrade call bolts broken) and the { last flexure breaks. The worst conceivable case involves the thermal shield dropping or moving downward eleven inches to rest'on the core barrel radial support keys. In addition to the analyses presented by the licensee, the licensee proposes to inspect the thermal shield during a June 1990 mid-cycle outa using the same. equipment and methods used during the January 1989.inspecge tion. The licensee also proposes to use two monitoring methods while in operation to detect any further degradation of the thermal shield: e neutron noise monitoring and loose-parts monitoring. The licensee
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proposes to shut down if it determines through the monitoring program that the. sixth flexure has isiled. 3.0 EVALUATION The video tapes recorded during the visual examination demonstrate that the thermal shield is still in its original position at this time. The tapes do not show any evidence of motion of the thermal shield. The tapes show that three bolts (two in one block; one in another) are protruding sufficiently far beyond allowable tolerances that it is reasonable to assume that they have drifted inward from vibration and are broken. Because the inspection was only visual it cannot be known if these are the only broken or cracked bolts. Licensee's vibration analysis concludes that the 240' and O' blocks are degraded and that the 300* block would probably degrade during cycle X. Furthermore, the licensee states and the inspection does demonstrate that the thermal shield is not in the " worst credible" condition or close to it because the last flexure is still intact and no support block wear or thermal shield motion has occurred. Support Llock wear would be expected if all of the bolts and dowel pins in r block were failed. No evidence exists to suggest that any individual support block assembly has progressed to this condition. The vibration analysis predicts that a third support b1cck will probably degrade during operation in cycle X, but that no damage will occur to the thermal shield. The analysis performed by the licensee used a simplified model consisting of beam elements and springs to represent linear and rotational stiffnesses of the system. The staff reviewed the pertinent information provided by the licensee and concluded that there are serious flaws in the methodology, modeling and in the evaluation of stresses which would result from the imoactive loads on the support blocks induced by the vibratory motion of the thermal shield. Because the analysis was found to be unacceptable by the staff, the license has been requested to l l
t . perform a mid-cycle inspection and to improve the proposed monitoring program whereby any further degradation of the shield supporting elements could be quickly detected and appropriate action taken by operating personnel. At the meeting with the licensee on May 1, 1989, the NRC staff presented its requirements for an insrtction of the thermal shield at the mid-cycle outage and changes to the proposed license conditions on thermal shield monitoring. The licensee agreed with the staff position and confirmed this agreement in its letters to NRC dated May 3 and 8, 1989. In evaluating the safety issues regarding the thermal shield the staff took under cemideration the following sequence of events which must take place prior to the situation which may cause a safety concern.. The scenario which would cause a concern is that the shielo could drop to the bottom of the reactor vessel and therefore obstruct the flow of . coolant to the core. In order that such a situation could exist the shield must be depriveo of its supporting elements and the following stages of further degradation would have to occur: (1 Failure of the sixth flexure (2 Failure of the all bolts at each support block (3 Shearing off the support blocks which hold the shield in the present position, and (4) Failure of the lower core radial supports The staff criteria require that the licensee provide an adequate neutron noise monitoring prograe which will detect any further degradation of any of the above elements and that the af ter failure of the remaining flexure, plant thuswill be shut down precluding immediately any further deterioration of the reactor internals. The staff believes that such an arrangement coupled with the mid-cycle inspection provides adequate assurance of safety. The noise signal from the ex-core power range neutron flux detectors will be recorded periodically and analyzed to monitor internal vibrations of the. thermal shield. Four accelerometers mounted on the reactor vessel flange will monitor acoustical noise in order to detect the possible appearance of loose parts in the lower dome of the vessel. The neutron noise analysis will probably not be effective in detecting gradual degradation of the fasteners, but failure of the last flexure would allow a large beam-mode oscillation at a much lower frequency which would be detectable. The occur, as discussed above. plant would be shut down for repairs should this The three bolts which were found to be broken will likely drift out all the way anc become loose parts at some point in cycle X. These parts will most likely fall to the bottom of the pressure vessel because of their weight and settle in a location of low flow velocity. Detection of , loose parts such as these under these circumstances would not be likely l l
due to the arrangement of the accelerometers which are mounted on the reactor vessel flange. In the unlikely event that the loose parts are. lifted up against the flow distribution or core support plates, no adverse impact is expected, and these impacts may be detectable since these locations communicate more directly with the accelero-meters. The licensee.and consultant (Westinghouse) have analyzed the changes in m etor aop 11"coolant flow barrel to the core to the radial core in the event the shield should tilt or supports. The changes in flow and flow distribution would be minor and within the design parameters, and. a'e therefore acceptable. We conclude that operation in cycle X as proposed is acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on May 11 1989 (54 FR 20459). Accordingly, based upon the environmental. assessme,nt, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that: (1) there is reasonable essurance that the health and safety of the public will not be encangered by operation in the proposed manner, l (2) such activities will be' conducted in compliance with the Commission's regulations and (3) the issuance of this amendment will not be inimical to the common defense ano security or to the health and safety of the public. Principal Contributors: R. Lipinski L. Lois ' C. Trammell N. Hum Dated: May 15, 1989
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, UNITED STATES NUCLEAR REGULATORY COMMISSION
. SOUTHERN CALIFORNIA EDISON COMPANY, ET AL.
DOCKET NO. 50-206 NOTICE OF ISSUANCE.0F AMENDNENT.TO PROVISIONAL OPERATING LICENSE The U.S. Nuclear Regulatory Comission (Comission) has issued Amendment No.127to Provisional Operating License No. DPR-13, issued to Southern California Edison Company and San Diego Gas and Electric Company (the licensees), for-operation of the San Onofre Nuclear Generating Station, Unit No.1, located in San Diego County, California. The amendment was effective as of the date of issuance. The amendment provides for a reactor vessel thermal shield monitoring pro-gram and mid-cycle inspection until the thermal shield fasteners are repaired during the. fuel cycle XI refueling and 10-year ASME Inservice Inspection. The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations. The Comission has made appropriate findings as required by the Act and the Commission's regulations in 10 CFR Chapter 1, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment and Opportunity for Hearing in connection with this action was published in the FEDERAL REGISTER on March 2,1989(54FR8854). No request for a hearing or petitions for leave to intervene were received. Subsequent to issuance of this notice, the licensees provided supplemental information by letters dated March 21 and 23 and May 3 and 8, 1989. These letters provided additional information and revised comitments encompassed by the original notice. i 9 0$WN 100
I q l The'Comission has prepared an Environmental Assessment related to this { action and has concluded that an environmental impact statement need not be prepared because operation of the facility in accordance with this amendment will have no significant. adverse effect on the quality of the human \ environment, l For further details with respect to the action see (1) the application f for amendment dated February 17, 1989, as supplemented March 21 and 23 and May 3 and'8, 1989, (2) Amendment No.127to License No. ' DPR-13, (3) the Comission's related Safety Evaluation and (4) the Comission's Environmental Assessment. All of these items are available for public inspection,at the Comission's Public Document Room, 2120 L Street NW., Washington,' DC20555, and at the General Library, University of Califor'ita, P.O. Box 19557, Irvine, California 92713. A copy of items (2), (3) and (4) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Director, Division of Reactor Projects - III, IV, Y and Special Projects. Dated at Rockville, Maryland this 15 day of May,1989. FOR THE NUCLEAR REGULATORY COMMISSION h, /r Charles M. Tramell, Senior Project Manager Project Directorate Y Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation i l
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Docket No. 50-206 Dr. Larry T. Papay, Senior Vice President Southern California Edison Company i Post Office Box 800 l 2244 Walnut Grove Avenue Rosemead, California 91770
Dear Dr. Papay:
. i
SUBJECT:
0PERATION OF SAN ONOFRE UNIT 1
- 1 This refers to the NRC Confirmatory Action Letter (CAL) dated January 31, 1989 and the CAL followup letter dated February 8,1989. These letters identified i' certain technical issues to be resolved before Unit I restart and requested yosr assessment of the aggregate significance of the various single failure and other design problems identified in recent months.
As discussed during our May 1, 1989 meeting and documented in the enclosed meeting report, your letters dated March 17 and April 18, 1989 presented the bases for your conclusion that Unit I could safely return to service. We concur with your characterization of the identified problems and recognize that you have established programs to identify other potential deficiencies. be We also by identified understand thatprograms these ongoing there well may(e.g.other design deficiencies enhanced engineering / design hereafter activities, design basis upgrade program) a,nd expect that they will be assessed and handled consistent with established procedures and the requirements of your license. Accordingly, the NRC hereby concurs with your intention to restart San Onofre Unit 1. This conclusion was reached in coordination with the Region Y Regional Administrator and is also based on your satisfactory response to NRC concerns, as documented in the meeting sumary and your written certification dated May 12, 1989 that all comitments made to the NRC for actions to be completed during the Cycle 10 refueling outage have in fact been completed. Sincerely, f 0> siq gg omasE.Mbs r pp - office of Nuciear Refor [eguiation
Enclosure:
Meeting Sumary cc w/ enclosure: See next page
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.! Mr. Kenneth P. Baskin San Onofre Nuclear Generating-Southern California Edison Company Station, Unit No. I cc Mr. Kenneth P. Baskin -
Vice President' Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770 David R.~Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 i Mr. Robert G. Lacy. Manager, Nuclear San Diego Gas & Electric Company ! P. O. Box 1831 San Diego, California 92112 Resident Inspector / San Onofre NPS U.S. NRC P. D. Box 4329- ' San Clemente, California 92672 Mayor City of San Clemente San Clemente, California 92672 -! Chairinan Board of Supervisors County of San Diego 1600 Pacific Highway Room 335 . San Diego, California 92101 L Regional Administrator, Region V i U.S. Nuclear Regulatory Comission ) l 1450 Maria Lane, Suite 210 t I Walnut Creek, California 94596 - 1 Mr. Paul Stalinski, Chief Radiological Health Branch State Department of Health Services 714 P Street, Office B1dg. #8 , Sacramento, California 96814 - 1
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. QUESTION 4 How much will repairs to the thermal heat shield cost, and how long will the' repairs take?. Will ratepayers pay the bill or will the shareholders pay part of the cost?
ANSWER
' Southern California Edison estimates that the cost to repair the thermal shield will be $10 to $15 million and that the repair will require 5 to 7 months to complete. Because the_NRC'does'not regulate rates, we are unable to answer your question regarding how the cost of the repair will be allocated. The method by which costs are recovered is determined by the California Public UtilitiesCommission-(CPUC).
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QUESTION 5 .If NRC allows San Onofre 1 to restart without repairing the y themal heat shield, how specifically and how frequently will Edison' perform monitoring 'of the' problem during the time the plant is allowed'to operate?..
' ANSWER p
The monitoring program is described in license condition 3.M. which-is part of 7 License : Amendment No. ,127. (attached to Question 3).-- i
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. j QUESTION 6.- Edison has said there is no safety problem even if the heat i j
. shield came loose and fell, because a " shelf" would catch it, and it would' fall only 11. inches. Isn't'there'a danger that if a j 24-ton. steel cyclinder fell 11 inches.that other bolts, supports-or other material might cause blockage of coolant and possibly I cause uneven or inhibited cooling? l ANSWER These matters.are discussed in the NRC staff's Safety Evaluation attached to License Amendment No. 127 (attached to Question 3).
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' QUESTION 7. When.was the most recent NRC " Systems Assessment of Lice'nsee Performance"'(SALPS)evaluationprocedureconductedforSan-Onofre 17 ANSWER' The'most recent Systematic Assessment of Licensee Performance (SALP) was .
issued on November 25, 1988, and covered the period October 1,-1987, through i September. 30, 1988. A copy of the SALP report is attached. i 4 0 1 I l
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.; . V ATTACHMENT TO QUESTION 7 f 91/31/1999' 16:00_ REACTOR SAFETY & PRO.T R5 ' 415 943 ,a(DD F.55-
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UNrrED STATES -
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- REGION V-
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WALNUT CRssE, CALIFORNIA esses W l'1'1999 Docket'Nos. 50-206, 50-361, 50-362
' .' ' Southern California Edison Company P.'O. Box 800 2244: Walnut Grove Avenue.
Rosemead, California 92770 Attention: 'Mr. David J. Fogarty Executive Vice President Gentlemen: This refers to the NRC's Systematic Assessment of Licensee Performance.(SALP)- Board Report dated November 25, 1988
-Station (SONGS); our management meeting of Decemberfor your. San Onofre Nuclear Gen 13, 1988,-
discussed the contents of.this report; and..the written comments provided induring which w your' January 13, 1989 letter in response to the.SALP Soard's report. During the December 13, 1988 meeting, we discussed your planned corrective actions in response to this assessment of your activities authorized by NRC - License Nos. OPR-13, NPF-10, and NPF-15. meeting are summarized in the enclosed meeting report. Subjects. discussed during Although fnur fianetional areas were considorod by tho SALP Bosed to be ~ 1 Category 1, your performance in the Engineering Assessment / Quality Verification areas was asses /sed Technical Support and to be Category 3. WeSafety-recognize the significant corrective actions your staff has undertaken in
~these areas, as discussed with you during the December 13 meeting. However, aggressive management attention is still needed to ensure that the defined corrective actions are fully and effectively implemented. .
As discussed with you during our November 2, 1988 management meeting, effective actions by your safety oversight groups will also be important to ensure that all engineering weaknesses are identified and included within the scope of your corrective actions. Your January 13, 1989 response described an independent review of oversight organization effectiveness which you are initiating with the assistance of a consultant. During following: the management meeting we also discussed NRC concerns regarding the Additional effort is needed by SCE management and by your Licensing and engineering staffs to improve the qua11ty of licensing submittals. We will monitor the effectiveness of improvements you are initiating in this ' area, as described in your January 13, 1989 letter. p a % M &- 3(R
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'Recent NRC initiatives have determined that, on an industry-wide basis, as much as half the risk of core melt is associated with periods of mid-loop-operation. Your management should ensure that special diligence is maintained by all plant personnel during such periods.
More effective measures are needed to address continuing discoveries of illegal drugs.inside the protected area of your facility. We have reviewed your January 13, 1989 . response to the SONGS SALP' Report and ccnclude that' the corrective. actions you described to effect improvement in identified areas of weakness are responsive and substantial.- We will discuss the progress of your corrective actions during future inspections and management meetings. Based on the formal exchange of information between our respective staffs, and in.the absence of verbal identification of discrepancies within the report or written comments from you which require resolution, no changes to the November 25, 1988 SALP Board Report are deemed necessary.- Related NRC conclusions are presented in Appendix'I to this letter. In accordance with 10 CFR 2.790(a), a copy'of this letter and the enclosures will be placed in the NRC Public Document Room. Should you have any questions concernin letter or Appendix I hereto, the issues discussed in the enclosed mee g repor , or the SALP Board's report, we will be pleased to discuss the. with you. Sin oly, _f $ J. B. Martin Regional Administrator
Enclosures:
- 1. Appendix I, NRC Conclusions.
.a.. . 2. SALP Meeting Report Nos. 50-206/88-32, 50-361/88-33 and 50-362/88-35
- 3. Licensee response to SALP Report, dated January 13, 1989
- 4. Final SALP Report Nos. 50-206/88-25, 50-361/88-26 and 50-362/88-28 cc w/ enclosures:
Dr. L. T. Papay, Senior Vice President C. B. McCarthy, Vice President (San Clemente) K. P. Baskin, Vice President H. E. Morgan, Station Manager (San Clemente) State of California 1
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APPENDD I E NRC CONCLUSIONS A. Comments Received From Licensee Southern California Edison's January 13, 1989 response to the San Onofre SALP of theBoard Report presented no comments addressing the specific content report. the SALP Report.Because there were nc comments,.no changes were made to B. NRC Conclusions Regardina Acceptability of Licensee's Planned Corrective . Actions k We to befeel that yourand responsive actions to deal with areas needing improvement appear substantial. i We. will review your progress during- .f our ongoing inspection program and as part of future management meetings. C. Recional Response Administrator's Conclusions Based On Consideration of Licen I have concluded changed. that the overall ratings in the affected areas have not. l 1 3 _ _ _ _ _ _ _ _ . _ . . _ _ - - - - - ^ - - - - ' ^ - '
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- m ewee v.es U.S. NUCLEAR REGULATORY COMMISSION REGION V Report Nos.
50-206/88-32, 50-361/88-33, 50-362/88-35 Docket Nos. 50-206, 50-361, 50-362 License Nos. OPR-13, NPF-10, NPF-15 Licensee: Southern California Edison Company P. O. Box 800, 2244 Walnut Grove Avenue Rosemead, California 92770 Facility Name: Fan Onofre Units 1, 2 and 3 Meeting Location: SCE Corporate Offices, Rosemead, California Meeting Date: December 13, 1988 Prepared by: h $det F. R.,//Huey, Senior Resident Inspector
'!3 U Date Signed San Ohofre Units 1, 2 and 3
. Approved By: 4WMA P.H./fchnson, Chief , 3U Date Signed ReactWr Projects Section 3 Meetino Summary Manacement Meetina on December 13, 1988 (Recort Nos. 60-206/88-32.
_50-361/88-33, and 50-362/88-35) A Systematic December 13, Assessment 1988 of License Performance (SALP) meeting was held on period October 1, 1987 through Septemberto 30, 1988.discuss the results of the most recen Other items of interest
. relating to the San Onofre Nuclear Generating Station were also discussed.
l l Sl W 50( i 1 _ _ _ _ _ _ _ _ _ . . i
g .. , o mm wm ene ew DETAILS
- 1. Meetino Participants I i
Nuclear Reculatory Commission (NRC) J. Martin, Regional Administrator D. Crutchfield, Acting Associate Director for' Projects, NRR { D. Kirsch Director Division of Reactor Safety and Projects G. Knighton, Director, Reactor Project Directorate V, NRR R. Zimmerman, Chief, Reactor Projects Branch G. Yuhas, BranchChief. Emergency Preparedness and Radiological Protection P. Johnson, Chief, Reactor Projects Section 3 C. Trammell, San Onofre Unit 1 Project Manager, NRR D. Hickman, San Onofre Units 2/3 Project Manager, NRR 1 F. Huey, Senior Resident Inspector J. Tatum, Resident Inspector A. Hon, Resident Inspector Southern California Edison company D. Fogarty., Executive Vice President
' C. McCarthy, Vice President, Site ManagerK. Baskin, Vice Pre R. Rosenblum, Manager of Quality Assurance D. Nunn, Manager of Nuclear Engineering & Construction ~
M. Medford, Manager of Nuclear Regulatory Affairs H. Morgan, Station Manager D. Heinicke, Deputy Station Manager D. Herbst, Quality Assurance Manager D. Stonacipher, Quality Control Manager R. Krieger, Operations Manage'- D. Shull, Maintenance Manager J. Reilly, Technical Manager P. Knapp, Health Physics Manager K. Slagle, Material & Administrative Services Manager O. Peacor, Emergency Preparedness Manager P. Eller, Security Manager J. Schramm, Operations Superintendent, Unit 1 V. Fisher, Operations Superintendent, Units 2/3 L. Cash, Maintenance Manager, Unit 1 R. Santosuosso, Maintenance Manager, Units 2/3 C. Chiu, Assistant Technical Manager M. Wharton, Assistant Technical Manager C. Couser, Compliance Engineer
- 2. Management Discussion Mr. Martin opened the meeting by stating that the primary challenge of the SALP process is to provide increased licensee and NRC attention on the actions which are needed for sustained excellent operation of
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licensedLfacilities. In this regard,'Mr. Martin noted that during this last.SALP period, the NRC and the.-licensee had focused significant atten-tion ared resource on the performance of engineering and technical work at i: " San Onofre. . This attention resulted in the identifiestion oficignificant .' deficiencies-(many of which were programmatic in nature) in the manner in
;which Onofre.engineering and technical; activities have been performed at San l Although SCE has initiated aggressive actions to correct the .
observed deficiencies. as noted during the management conference. held at ' i
.e -the site on November 2, 1988. the low SALP marks in the Engineering / Tech -
'nical Support and Safety. Assessment / Quality Verification ' areas reflect
'the significance of these deficiencies.and the shortcomings of management -
, involvement and quality. oversight which allowed them to remain so long' undetected.
Mr. Martin encouraged SCE management to.not only follow through with identified corrective. actions, but to apply. the lessons . learned from the engineering program review to other areas of probable benefit. In particular Mr. Martin' noted that additional attention will be placed on licenset maintenance activities during the new SALP period.
.Mr. Kirsch.briefly In particular, ha ;noted:
reviewed the results of the November 25 SALP report. 1. Significant strengths were noted in' operations .(particularly in the areas of trip reduction.. operator knowledge and well-written proce-dures) and Security in the Radiation Protection, Emergency Preparedness'and-areas. 2.. ~ Failure to follow procedures was noted as a continuing problem in the Maintenance / Surveillance area. 3. 1 The.most significant problems were noted in the Engineering / Technical Support and Safety Assessment / Quality Verification areas,- although significant improvement was noted in root cause evaluation and chemistry program implementation. In particular, major weaknesses contributing to the assignment of a Category 3 rating in these areas were observed to include: Enoineerina/ Technical ~Secoort Insufficient understanding of plant design Inadequate control of design processes Inadequate design data base Safety Assessment /Ouality Verification
- Insufficient management involvement /self critical attitude Ineffective quality oversight group involvement Inadequate safety reviews Improper deportability determinations Inadequately supported amendment requests Mr.
tals.Crutchfield briefly reviewed NRR concerns with SCE Ifeensing submit-In particular, he noted that Unit I submittals had not been of the
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{ 3 same quality as Unit 2/3 submittals. { In this regard, he noted that SCE needed to place more attention on the quality of all Unit I submittals { and not just those needed to support continued plant operation. Mr. i also a significant concern.Knighton noted that late and ' inadequate ! In this regard, Mr. Martin added that he { believed that the problem may involve a tendency.on the part of the licensee the to not tell the NRC about developing problems until they have full story. )
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timely with and the NRC. open discussion of potential or significant plant p j that SCE would strive for improved communications.l.icensee ma .{ j Mr. Baskin stated that SCE would like to briefly review the specific i corrective actions which are being pursued to improve performance in the ( Engineering / Technical Support and Safety Assessment / Quality Verificatio{ functional areas. : Mr. nicalNunn Supportdiscussed area. specific actions being taken in the Engineering / Tec ; With regard to staffing, he noted that the existing 1evel of aboutof150 engineering engineers. personnel was about 119, with a near-term project{ Mr. Martin questioned the ratio of engineering. supervision to working engineers, noting thet other su supervision requiring about 33% of the total engineering effort. Mr. Nunn. stated that SCE was sensitive to this concern and would put emphasis on effective engineering supervision. Mr. Rosenblum and Mr. Medford discussed the actions being taken to improve performance involving quality oversight groups and licensing activities. Mr. Knapp on the actions being implemented by SC which began in late November. spent fuel particle contamination during In closing improve per,formance in the above areas and stated that
, , - significant improvements would result from SCE's efforts.
He noted that it is often easier performance. to correct bad performance than to sustain excellent to maintain its level of effort in all areas impacting safe plan operation in order to provide continued excellence in this area. Mr. Martinhe In particular, then notedled thatthe discussion recent to other NRC initiatives associated areas withof current NR operation at reduced coolant inventory (i.e., mid-loop operation of plant shutdown periods. shutdown cooling) had underscored the significani In this regard, Mr. Martin stated that licensees should take prompt actions to dispel any lingering attitudes by plant operating periods. personnel that diligence can be relaxed during plant shutdown as half of He the noted that total risk of recent core melt NRR reviews have is associated with established plant shutdown that as much periods. He requested that SCE management provide specific attention to I
C1/31/1989. 19:04 REACTOR SAFETY 8. PROJ R5 415 943 3755 P.12 I . L 4
. periods of mid-loop plant operation. ensure that special diligence i
{ Mr. Martin noted that SCE was continuing to find illegal drugs within the protected areas of the plant. He emphasized that continued poor perfor-mance in this area could not only undermine NRC confidence, but might also affect public ennfidence in the adequacy of the licensee's drug program. the problem was the continuing finds in the Unit 2/3 lube He oil roo stated that SCE must quickly implement effective measures to stop this problem. considered.Mr. 51agle briefly discussed measures being taken or
.as . .
ffa nah,, ^ . UNtTED STATES
! c
. p.' .- [ I. NUCLEAR REGULATORY COMMISSION nEGION V N- *,
!* 1450 MARIA LANE. SUITE 210
%, . . . . . / WALNUT CREEK, CAUFORNIA 940065388 NOV 2 5 Igg Docket No. 50-206, 361, 362
' Southern California Edison Company P..O. Box 800
- 2244 Walnut Grove Avenue Rosemead, Ca11fornia 92770 -
Attention: Mr. Kenneth P. Baskin, Vice Pre'sident Nuclear Engineering,' Safety, and Licensing Gentlemen:
Subject:
Systematic Assessment of Licensee Performance (SALP) The NRC Systematic Assessment of Licensee Performance (SALP) Board has completed its periodic evaluation of the perfomance of your_ San Onofre Nuclear Generating Station during the period October 1,1987 through September 30, 1988. The performance of San Onofre was evaluated in the functional areas of plant operations, radiological controls, maintenance / surveillance, emergency prepa rednes s , , security, . engineering /technica l support, and safety assessment / quality verification. The criteria used in conducting this assessment and the SALP Board's evaluation of your performance in these functional areas are contained in the enclosed SALP report. Based upon discussions with your staff, a management meeting to discuss the results'of. 1988, the SALP at your Rosemead Board's assessment has been scheduled for December 13; offices. be discussed further with your staff in the near future. Arrangements for the manag Overall, the SALP Board found your perfomance to be acceptable and directed toward safe facility operation. Good perfomance was noted to have continued in the Plant Operations area; and Radiological Controls Emergency Preparedness, and Security were also perceived as areas of strength. However, several concerns were identified by the NRC during this assessment period. The Board found that there was a need to improve the involvement of management in developing a self-critical attitude at all levels of your organization. This need was common to several functional areas, and most pronounced in the. areas of engineering and safety assessment. Engineering / Technical Support - insufficient understanding of plant design, design data' inadequate base. control of design processes, and discrepancies in the Safety Assessment / Quality Verification - insufficient aggressiveness of quality groups in identifying developing plant problems, inadequate safety reviews, improper deportability determinations, and inadequately supported amendment requests. fi ITT[51W n ee -
p. fe <
' JCV !. 5 1968- .
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' Actions were. initiated late in the assessment period to start improvements -
but these actions are still in the formative stages..
. Aenclosed management report.summary of.this assessment is provided-in Section II of the Perceived strengths and weaknesses and Board recommenda-tions are discussed in Section IV, Performance Analysis. ..
l You are requested to provide to this office, within 30 day: after our ganage-ment meeting, a written response which addressas the two functional areas !" assessed by the SALP Board as Category 3. This response should describe-actions which.you have taken or plan to take to provide improved performance in these. functional ~ areas. Actions described in previous correspondence may be included by. reference if appropriate. Your response ma on or amplification of.the SALP report in other areas, yas also include comments appropriate. In accordance with Section 2.790 of the NRC's " Rules lof. Practice," Part 2, Title. IO, Code of Federal Regulations, a copy of this letter, the enclosed SALP
. report, and your response will be placed in the NRC's Public Document Room.
lhe NRC's Offica for Analysis and Evaluation of Operational Data performed an-assessment of licensee event reports submitted for San Onofre. .This assess- , ment was provided es an. input to the SALP process; a ucopy is- therefore provided-as Attachment 1 to the enclosed report. o - The' response requested by this letter is not subject to the clearance proce-dures of the Office of Management and Budget as required by.the Paperwork Reduction Act of 1980 PL 96-511. Should you have any questions concerning the SALP report, we will be pleased to discuss them with you.
- nc e1y, e
-J. B. Martin Regional Administrator
Enclosure:
SALP Report No. 50-206/88-25, 36 jo8-26, 362/88-28 Attachment 1 Enclosed in SALP Report cc w/ enclosure: D. H. Fogarty, (Rosemead) C.'B. McCarthy, (San Clemente) H. E. Norgan, SONGS (San Clemente) State of California o _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ - - . - . _. .)
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SALP BOARD REPORT
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U. S.-NUCLEAR REGULATORY COMMISSION REGION Y SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE 50-206/88-25, 361/88-26, 362/88-28
. SOUTHERN CALIFORNIA EDISON COMPANY
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SA DNOFRE NUCLEhR GENERATING STATION OCTOBER'1, 1987 THROUGH SEPTEMBER 30, 1988 6 O e P u
/ , y 8-TABLE OF CONTENTS ' Page.
.I.
Introduction . . . .~. . . . . . . . . . . . . . . . ... . .. 1
'A.
B. ^Direct Licensee Activities . . . . . . . . . ... . . . . . . . Inspection and Review Activities . .. . . . .. . . 1 4i L- II. Summary of Results . . . . . . . . ... . . . . . . . . . . .. 4. l '. l ' A. Effectiveness of Licensee Management. . . . . . . . . . B. 4 Results of Board. Assessment . . . . . . . 5 C. Changes in SALP Ratings . . . . . . . ........ ........ 6 I I I . C ri te ri a . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 I V. . Pe rfo rma nce Ana lys i s . . . . . . . . . . . . . . . . . . . . . 7-A.- Plant Operations. . . ... ... . . . . . . . . . . . . 8 Radiological Controls . . . . . . . B. C. . . . . . . . . 10 Maintenance / Surveillance. . . . . . .
........ 13 D. - Emergency Preparedness. . . . . . . . 15 E.- Security. . . . . ....... . . . . ... . . . . . . . . .
......... 17 F. Engineering / Technical Support . . . ... . . . . . . . . 19
. G. Safety Assessment / Quality verification. . . . . ... . . 21 Y. Supporting Data and Summaries. . . . . . . . . . . . . . . . 25-A. Enforcement Activity ................. 25.
B. Confirmation of Action Letters ............ 25 C. Other
......................... 26 ..
TABLES-Table 1 - Inspection Activities and Enforcement lSummary Table 2 - Enforcement Items - Table.3 - Synopsis of Licensee Event Reports _,___--.--uu. .
e
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I. INTRODUCTION The. Systematic Assessment o'f Licensee Perfomance (SALP) is an NRC staff integrated effort to collect available observations and data on a periodic basis and evaluate licensee's performance based on this infomation. The program is: supplemental. to normal regulatory processes used to ensure compliance with NRC rules and. regulations. It is intended to be sufficiently diagnostic to provide a rational basis for allocating NRC' resources and to provide meaningful feedback to the licensee's management regarding the NRC's assessment of their facility's performance in each-functional area.
- An NRC SALP Board, composed of the members listed below, met in the Region V office on November 9,1988, to review observations and data on the licensee's performance in accordance with NRC Manual Chapter 0516, n " Systematic Assessment of Licensee Performance " dated June 6,1988. The
. Board's findings and recommendations were forwarded to the NRC Regional Administrator for approval and issuance.
This report is the NRC's assessment of the licensee's safety performance at San .Onofre for the period October 1,1987 through September 30 1988.. The SALP Board for San Onofre was composed of:
**D. F. . Kirsch,. Director, Division of Reactor Sa fety and Projects.
- Region V (Board Chairman)
**R. A. Scarano, Director, Division of Radiation Safety and Safeguards
**G. W. Knighton, Director Project Directorate Y. NRR
**R. P. Zimerman, Chief Reactor Projects Branch
*G. P. Yuhas, Chief, Emergency Preparedness and RadiologicO Protection Branch
**P. H. Johnson, Chief, Reactor Projects Section 3 **
*H. S. North Acting Chief, Facilities Radiological Protection Section .
*M. D. Schuster, Chief, Safeguards Section
**C. M. Trammell, Unit 1 NRR Project Manager
**D. E. Hickman, Units 2 and 3 NRR Project Manager
**F. R. Huey, Senior Resident Inspector
**C. W. Caldwell, Project Inspector
*J. E. Russell, Radiation Specialist
*G. M. Good, Emergency Preparedness Analyst
*D. W. Schaefer, Safeguards Inspector
- Denotes voting member in functional area of co
** Denotes voting member in all functional areas.gnizance.
A. Licensee Activities In general, all three units operated satisfactorily dur ing the assessment period and were relatively free of problems. Specific operational events were as follows: - __ - - _ - - - --- .__._______,,---:-----=--"--"---~"'~'- '
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2 _ ,, Unit 1 Unit 1 operated essenticily at full. power from the beginning of the assessment period until mid-February 1988. The Unit shut down on February refueling). 13,- Durin 1988 for a planned 45-day maintenance outage (no qualification (EQ)g that outage, problems with environmental Unit. The residentofinspectors components became a major issue concerning the and the licensee identified se.veral safety related electrical components
. qualified. These problems were-indicative that were not properly' of a progranrnatic breakdown of design controls associated with the licensee's'EQ program. The licensee initiated a comprehensive reevaluation of the EQ program which identified more than 140 additional components which were not properly included in the program. The root cause of the EQ program breakdown was determined to be inadequate design controls during the period between 1981 and 1984, and an inadequate review of electrical interactions, as required by 10 CFR 50.49 (b)(2). As a
- result, the licensee delayed the startup from the mid-cycle outage and instituted a ceaprehensive program to identify and correct all EQ deficiencies prior to restart.
Another problem developed on May 31, 1988 (during the outage) which concerned the capacity of the emergency diesel generators (D/Gs). The licensee found that the design calculations for the Unit 1 D/Gs did not have sufficient capacity to handle all post-accident loads
' due to D/G derating which occurred in November 1985. For corrective
' action, SCE c5tained a Technical Specification change to increase the allowed losu (effective until the next refueling outage). For long .'
term corrective action, the licensee plans to replace necessary parts in accordance with vendor recommendations so that the capacity of the D/Gs may be raised back to the nameplate value of 6000 Kw. These problems were resolved by the licensee and the Unit was ' restarted on August 5,1988 after wnpletion of the 174 day mid-cycle outage. The outage was extended 130 days to resolve the EQ issues discussed above. The Unit operated at full power through the remainder of the assessment period. Unit 2 Unit 2 was in a refueling outage at the beginning of the assessment period. The outage was free of any significant problems and the Unit was restarted on December 9, 1987. Other than a manus 1 trip due to the failure of a feedwater isolation valve in mid-December 1987, the Unit operated at power until March 16, 1988 wNn ' t was shut down as i a result of steam generatcr tube leakage. Tla scace of the leakage i was a previously plugged tube from which the A had fallen. This plug was replaced, others were inspected, and the plant was restarted on April 4, 1988. The Unit operated at essentially full power until May 6, when the licensee initiated a shutdown (per Technical ! Specification 3.0.3) as a result of both emergency chilled water (ECW) system chillers being declared inoperable due to low Freon 1 level. The problem was corrected and the power decrease was terminated after about three hours. 1
E L. ,. , 3 ,~ The Unit resumed full power operation on May 6,1988 and operated. continuously until-August 21, when an Unusual Event (UE) was declared. and a shutdown was initiated due to an actuating relief valve on one of the four safety injection tanks (SITS). SCE terminated the UE after completion of the controlled reactor shutdown. The licensee corrected the source of the problem, which was a roughly machined surface between the valve stem and the stem guide. Similar. corrective action was taken for one other SIT relief valve (Unit 3. SIT relief valves were inspected and found to be acceptable). The-Unit was subsequently restarted on August 23, and operatad at full power for the remainder of the SALP period.
- Unit 3 The Unit operated at full power at the beginning'of the SALP period until a reactor trip occurred on October.11 due to influx of seaweed into the main condenser. The Unit was restarted the next day and operated at full power until January 23, 1988, when the Unit was shut down for 16 days due to a main generator hydrogen leak. Except for a manual trip on February 20, prompted by a spurious engineered safety features actuation, the Unit operated at' full power until April 30, 1988, when the licensee shut it down to begin the Cycle 4 refueling outage.
On June 22, 1980, with the Unit shut down, approximately one foot of
- water was inadvertently siphoned from the spent fuel pool to the
- reactor cavity due to failure (during initial plant construction) to install a vacuum breaker in the purification system piping which.
extends to the bottom of the fuel pool. .A second event occurred on
' June 23, while licensee personnel were preparing to transfer water from the reactor cavity to the spent fuel pool, because personnel left a temporary pump unattended in a primed condition. For corrective action, the licensee instituted precautions and cont'rols to prevent siphon paths. For the long tenn, a design modification was planned to install vacuun. breakers in Unit 2/3 spent fuel pool -i purification suction piping as originally specified in the FSAR. j On July 7, 1988, during draindown of the reactor vessel, cavitation of the operating low pressure safety injection (shutdowa cooling) pump occurred on two occasions due to blocking of a reference level sensing port (this caused the reactor vessel level .inhcation te read incorrectly). The draindown was terminated until the problem was identified. Operator attentiveness was credited for avoiding a potentially serious problem, (a loss of shutdown cooling condition) although the event identified a need for improved control of maintenance activities.
Unit 3 was restarted on August 16, 1988 after completion of the 3-1/2 moath maintenance outage. The restart had been delayed approximately one month to complete repcirs to a shutdown cooling isolation valve and replace seals on a reactor coolant pump. The Unit was subsequently shut down on August 26 to correct unisolable tube Irakage in a fifth point feedwater heater. The Unit was returned to service on August 29, after repair of the heater, and operated at full power for the remainder of the assessment period. __ m.mmamm
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4 B. Direct Inspection and Review Activities Approximately 5480 on-site inspection hours were spent in performing a total of 36 inspections by resident, region-based, headquarters, and contract personnel. Inspection activity in each functional area is sumarized in Table I. II. SUMMAR OF RESULTS . A. - Effectiveness ~of licensee Management .
. Notable licensee achievements wre observed during this SALP period, including a significant reduction in the number of reactor trips and relatively low forced outage rates of 7% and 5% for Units 2 and 3, respectively. Plant performance included a number of notable strengths. However, several weaknesses were also observed during the assessment period. The most significant of these weaknesses concerned engineering and technical support activities, licensing activities, and a lack of aggressiveness of safety oversight groups in identifying engineering / technical deficiencies.
The performance of the Plant Operations staff was very effective during this period, with strengths observed in staffing and professionalism. The alertness of control room operators was - credited on one occasion with averting a potential loss of shutdown cooling flow caused by poor control of maintenance activities. The ifcensee also demonstrated an aggressive radiological controls program which served as an industry leader in several respects. Effective management controls, ample and capable staffing, and self-critical attitudes also provided good overall performance in the Emergency Preparedness and Security areas. The Board considered the licensee to have an effective Mcintenan'ce and Surveillance program, although weaknesses were observed in the control of maintenance activities and in compliance with maintenance procedures and instructions. Weaknesses were also observed in the Engineering / Technical Support functional area. The licensee was found to have a depth of personnel and material resources in this area, and performed many program requirements in an effective manner. However, a number of significant engineering and technical problems (discussed in Sectics IV.F) were manifested during this SALP period which reflected adversely on the quality of engineering work and the effectiveness of the administrative controls which govern it. While it is true that some of the problems were identified by more aggressive engineering or quality verification performance, and actually resulted from poor engineering work during prior SALP periods, a need for improved engineering / technical performance was clearly indicated. Also apparent was a need to improve the completeness and correctness of the plant's design basis documentation.
[ %, U.o _ l 1 Other asse.ts in the Safety Assessment / Qual'ity Verification functional j area included-an improved root cause assessmant program and an i effective- program for monitoring plant perfomance. However, several significant weakness were noted in program implementation in this functional area.. In particular, the quality assurance organization i and the quality oversight groups showed insufficient aggressiveness in identifying problems in the plant engineering and technical support, area, and in the identification of significant safety. issues in general. . In addition, in a number of cases, the licensee's :j a timeliness and adequacy of.11 censing tubmittals and timeliness of ^ deportability evaluations were inadequate. The weaknesses noted above were discussed during periodic meetings 1 with licensee. management. These discussions emphasized a need for a self-critical attitude by SCE in addressing areas of weakness, particularly during the early portion of the SALP period. In a manner indicative'of such a self-critical attitude, senior SCE management recognized the significance of the observed weaknesses in the Engineering / Technical Support'and Safety Assessment / Quality
- Verification areas and initiated comprehensive actions late in the SALP period to provide improvement in these areas. These involved a corporate reorganization to put all such activities under one Vice President, plans to move the department closer to the San Onofre Station, and a review and updating of the plant's design basis documents. These actions, if vigorously pursued, should significantly improve the quality of engineering and safety
, assessment programs which support San Onofre. ,
B. Results of Board Assessment Overall, the SALP Board found the perfonnance of NRC licensed activities by the licensee to be acceptable and directed toward safe operation of the San Onofre Station. The SALP Board has made specific recommendations in most functional areas for licensee management consideration. The results-of the Board's assessment of the licensee's perfonnance in each functional area, including the previous assessments, are as follows: Rating Rating Last This Functional Area Period
- Period Trend **
A. Plant Operations 1 1 None B. Radiological Controls 2 1 None C. Maintenance / 2 2 None Surveillance D. Emergency Preparedness 1 1 None E. Security 2 1 None F. Engineering / Technical 2 3 None Support : G. Safety Assessment / 2 3 None Quality Verification No trend was apparent for any of the functional areas during this period.
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- I Maintenance and Surveillance were separate functional areas during the last SALP period. However, both areas received a-rating of 2 during the last assessment. Safety Assessment /
Quality Verification is a new functional area this period. It is similar to, but more comprehensive than, the Quality Programs and Administrative Controls. Affecting Safety functional area
, which it rep ~ laced. -
k Other functional areas rated separately '
- during the last SALP period, such as Fire Protection and Training, were evaluated as appropriate within the scope.of the functionaljreaslistedabove. ,
The. trend indicates the SALP Board's appraisal of the licensee's direction of perfonnance in a functional area near the close of the assessment period such that continuation of this trend may result in a change in performance level. Determination of the
.. performance trend is made selectively and is reserved for those instances when it is necessary to focus NRC and licensee-attention on an area with a declining performance trend, or to acknowledge an improving. trend in licensee performance. It is -
not necessarily a comparison of performance during the current period with that in the previous period. C. Chances in SALP Ratinos ., The licensee's overall performance was observed to have improved in the Radiological Controls and Security areas during the period due to the strong performance exhibited by these organizations, as discussed in Paragraphs IV.B and IV.E. 'The licensee's performance in the Engineering / Technical Support area declined frem Category 2 to Category 3 during this period, based primarily upon a number of significant engineering problems which were observed by the licensee and the NRC during the period, as discussed further in Paragraph IV.F. Performance in the Safety Assessment / Quality Verification funct'ional area also declined inadequately su from Category 2 to Category 3, due primarily to detenninations,pported licensing submittals, improper deportability
. and a perceived lack of aggressiveness by quality oversight groups in identifying problems with engineering / technical activities, as discussed in. Paragraph IV.G.
III. CRITERIA Licensee performance is assessed in selected functional areas, depending on whether the facility is in a construction or operational phase. Functional and the environment. areas normally represent areas significant to nuclear safety . I Some functional areas may not be assessed because of little or no licensee activitSs or lack of meaningful observations. Special areas may be added to highlight significant observations. The following evaluation criteria were used, as applicable, to assess each functional area: 1. Assurance of quality, including management involvement and control. e
(, , , i '. ,
>, 7.
2. Approach to resolution of technical issues from a safety standpoint.
- 3. - Responsiveness to NRC initiatives.
- 4. ' Enforcement history.
5. Operational events (including response to, analysis of, reporting of, and corrective actions for events). . i
- 6. Staffing (includingmanagement).
7. Effectiveness of the training and qualifications program. However, used wherethe NRC is not limited to these criteria and others may have been appropriate. On the basis of the NRC assessment, each functional area evaluated was rated according to three performance categories. The definitions of there-performance categories are as follows: Category 1: Licensee management attention and involvement are readily evident and place emphasis on superior performance of nuclear safety or safeguards activities, with the resulting performance substa'ntially exc~ee' ding regulatory requirements. Licensee resources are ample and effectively used so that a high level of plant and personnel performance is being achieved. Reduced NRC attention may be appropriate.
.Cateoory 2: Licensee management attention to and involvement in-the performance of nuclear safety or safeguards activities are good. The licensee has attained a level of performance above that needed to meet regulatory requirements.
Licensee resources are adequat,e pnd reasonably allocated so that being achieved. NRC attention maygood plant and personnel performance is be maintained at normal levels. 3
- Category 3: Licensee management attention to and involvement in the performance of nuclear safety or safeguards activities are not sufficier.c. The licensee's. performance does not significantly exceed that needed to meet minimal regulatory requirements. Licensee resources appear to be strained or not effectively used. NRC ;
attention should be increased above normal levels. IV. PERFORMANCE ANALYSIS The following is the Board's assessment of the licensee's performance in l each of the functional areas, plus the Board's conclusions for each area and its recommendations with respect to licensee actions and management emphasis. i i i
y ( f.'*,3 , a < 8: A.; Plan't Operations I'. . Analysis During the SALP period, approximately 1800 hours of direct inspection effort were applied.in;the Plant Operations area. Plant Operations continued to be.a licensee strength. The. licensee was noted to have had several significant- '
. accomplishments in the operations area during this SALP period.
The most significant was the reduction in the number of* reactor' trips. Other strengths were also observed regarding operator-knowledge and the adequacy of procedures. The primary areas in which improvement appeared warranted involved enhancement of' control over.the work authorization process and improved interface among the operating, maintenance, and technica1 organizations.. The resident inspectors observed licensee operation of the units on a daily basis, including random backshift hours. Operations
% staffing was observed to be adequate and control room operators -
were consistently. observed to be knowledgeable, attentive to plant conditions, and professional in their conduct. One example of exemplary performance was the prompt recognition and mitigation of an incipient loss of shutdown cooling'during..the.
- Unit 3 'efueling r outage when the reactor coolant system was
- being drained to mid-loop. Although this indicated a weakness in Operations control of maintenance activities ~, the alertness of the control room operators was credited with preventing a potential loss of ~ shutdown cooling flow. 'This event is '
discussed further under Maintenance / Surveillance, Section IV.C.. The licensee's approach _ to the resolution of operational safety issues was generally sound. The ifcensee's Trip Reduction Program,-initiated in~1g86, has been effective in achieving a goal of not more than one unplanned reactor trip per reactor year. Performance improved significantly during this SALP' period (a total of 3 trips this period compared.to 16 trips last- ! period). Unit I experienced no reactor trips during 190 days of power operation. Unit 2 experienced one manual trip (due to failure of a feedwater isolation valve) during 268. days of power operation. and Unit 3 experienced one automatic trip (10w condenser vacuum due to influx of seaweed) and one manual trip (prompted by a spurious ESF actuation) during 235 days of power operation. ' A sense of conservatism was generally exhibited by the Operations Staff when dealing with safety significant problems. 4 { A specific exception involved improper followup and operability determinations following observed low Freon levels on Unit 2/3 emergency ch111ers. The low Freon level was not properly understood or corrected for approximately one month, eventually contributing to inoperabf11ty of both emergency chillers. This ] indicated a weakness in interface among operations, maintenance, and technical organizations, since the plant
9 P operators had ample opportunities to resolve questions With cognizant station technical personnel. Inspection activities during the SALP period identified one Severity area. Level IV violation associated with the Plant Operations This involved failure to comply with a Unit 1 procedure for maintaining backup the operability of the auxiliary feedwater (AFW) nitrogen system. During this SALP period, a total of ten LERs were issued in the Plant Operations area. For Unit 1, three LERs were is' sued during the period. Of these, two were.due to equipment failure and one was the result of an inadequate procedure. Five Plant Operations LERs were issued for Unit 2. Of these, two were the result of operator error and three. were the result of equipment failures. Two operations related LERs were issued for Unit 3; one was for a plant trip due to low condenser vacuum following an excessive influx of seaweed, and the other concerned an inadvertent containment purge isolation syste' 'CPIS) actuation due to inadequate communication between operations and health physics personnel. On-line performance for the three units declined slightly during the 365 day SALP period compared to the last SALP period. However, this was largely due to licensee corrective actions resulting from Unit 1 EQ design problems. During the period, Units 1, 2, and 3 had unplanned outage rates of 36% (up from 9% 1ast period), 7%, and 5%, respectively. It is noteworthy th." 3 none of the trips or unplanned octages resulted from operator error. The licensed operator training program was characterized by excellent performance during the SALP period. This was *
- evidenced by a high pass rate of 92 percent (22 passes of 24 candidates)onreplacementexaminations. The facility also received a satisfactory evaluation for the Units 2/3 Requalification program from a pilot Requalification Program Evaluation conducted under a proposed change to Examiners Standard ES-601, "Requalification Program Evaluation". The facility expended a large amount of manpower and produced a quality product for its voluntary participation in this pilot avaluation. Their efforts included preparing job perfomance measurements, simulator scenarios, and a two-part written examination. The preparation of this material involved many changes from prior practice and required the production of entirely new material. The licensee had an acceptable pass rate of 86 percent (10 passes of 12 examinees) for this 1 i
Requalification Program Evaluation. l The Board concluded that the licensee's approach to plant operation was generally conservative and safety conscious. There was consistent evidence of prior planning and assignment of priorities. Briefings ("tailboard meetings") were observed to be conducted with involved personnel prior to plant l l
. ,f ,
10 Y evolutions.and testing. A specific strength was observed concerning' operating procedures, which were noted to be consistently well written, understood and implemented. . Decision-making'was usually at a level that ensured adequate review. _ An exception was the licensee's improper use of Special Orders as. interi:n emergency procedures for handling postulated ESF single failure events. The licensee took prompt corrective action when this deficiency was pointed out by the resident inspectors. In this and other cases, interface by the NRC generally showed the various levels of licensee management to be professional and responsive. In addition, plant housekeeping conditions were observed to be improving.
- 2. Conclusion Performance Assessment - Category 1.
- 3. Board Recommendations The Board recommends that the licensee continue management emphasis on trip reduction, prot.edure compliance, ettention to detail by equipment operators, and housekeeping. Action should also be taken-to strengthen the interface among Operations, Maintenance, and Technical personnel in a manner which will
- provide a more conservative approach to the resolution of plant problems.
B. Radiological Controls
- 1. Analysis
- This functional area was reviewed routinely during the .
assessment period by both regional and resident inspectien* staff. Over 620 hours of direct inspection effort were expended in this area. Strengths identified included a comprehensive management control system, a highly qualified staff, a fully accredited training and qualification program, and a comitment at the highest levels of management to quality performance. Housekeeping was effective, and contaminated areas were minimized. Observed weaknesses included minor deficiencies involving the implementation of a quality assurance program for auditing the use of packages of greater than type A quantities of radioactive material, the posting of a radiation and a high radiation area, and the failure of a maintenance worker to follow Health Physics (HP) requirements which resulted in an exposure in excess of the quarterly whole body limit. None of these problems appeared to indicate any programmatic weakness in radiological controls. The management control system was considered a strength in the Radiological Controls area. The HP division instituted a specific organization, during this period, to assure prior planning and assignment of priorities to the HP aspects of outage work. Hp policies were well stated and disseminated _ _ , _ _ _ _ . _ _ _ . - - - - - - - -_ - - - - - - - ~
o b .' ,a7 - *; 11 through routine staff meetings, a monthly newsletter. and monthly luncheons at which the HP Manager directly interacted with the line staff. Corporate management was frequently and effectively involved in site sctivities and performed monthly audits of specific aspects of the HP program. Corrective actions for identified deficiencies were typically effective and i the licensee was responsive to expressed NRC concerns (e.g., the licensee's efforts to deal with radioactive gaseous effluents . which were in excess of.the national average). Management review of HP problems has been addressed by an Operational Excellence Forum, which included all site managers. A' 1 management tour program was. instituted this assessment period which assured that all site management performed weekly inspections of ongoing work. The staff was also considered a strength in the HP area. Positions were well defined and authorities and responsibilities were clearly delineated. The staff was-highly qualified technically, with six certified health physicists on-site and one at the corporate office. Professional industry activities were supported monetarily and encouraged by management. Experience levels of personnel were high and the turnover. rate was low. During the period, the staff demonstrated a clear understanding of technical issues, notably in their implementation of an industry benchmark hot particle control program. In addition, conservatism was generally exhibited in problem resolution. Three violations wert identified during this assessment period, as indicated in Table 2. Most were isolsted occurrences which did not indicate any programmatic deficiency, and all were expeditiously and comprehensively corrected. During this SALP period there were few significant operational HP events,"but
-there were numerous monitor failures and spurious engineered safety features (ESF) actuations. These events were promptly and adequately reported. However, technical resolution of these events was slow. Also of concern was the fact that the licensee has been slow to complete the program for validation, verification, and documentation of safety affecting software in the HP area.
The licensee's training program has been fully accredited by the Institute for Nuclear Power Operations and was considered a strength. The instructors were primarily National Registry of Radiological Protection Technologists (NRRPT) certified, and were found to have implemented a well defined program of routine, job specific, and mock-up training. A complete program for contract technician training and qualification was also implemented which required satisfactory completion prior to the conduct of work. All SCE HP technicians were American National Standards Institute (ANSI) qualified with the exception of one person. In addition, management encouraged and supported training of technicians to become NRRPT certified. A program for feedback was also established to provide input of i
7,e', , 12 operational problems and concerns to the Training Department for use in periodic retrai6ing of personnel.- Procedures and policies were clearly defined and followed. In the few instances where policies were'r subsequently identified, there were followed and deficiencies no indications that were inadequate training was the cause . Another strength in this functional area was the licensee's demonstrated commitment to quality performance. -The site instituted a Performance, Recognition, Innovation, Dedication, Excellence (PRIDE) program to reward and recognize employees and-e groups which contributed to the achievement of goals in, among-others, exposure. the reduction of radioactive waste and occupational There was also a Productivity Improvement Program (PIP) which recognized and rewarded management and Operations
. personnel for exceptional contributions to quality service
. specifically in the area of limiting personnel exposure and improving access control to radiological areas.- In addition to these site-wide programs, there were internal HP incentive ,
programs to acknowledge exceptional contributions by line personnel (The Silver Dollar Program) and for. contributions in the area of dose minimization (ALARA awards). The Quality Assurance organization also demonstrated expertise in the HP area and provided independent critical review of the program, particularly in the area of radioactive material
- control and the Radiation Exposure Permit program. The licensee took exceptional efforts to deal with the root cause of the hot particle and elevated gaseous effluent problems discussed previously by performing audits of their fuel supplier's ,
' fabrication facilities in order to minimize or eliminate fuel integrity problems. The licensee also took the lead in obtaining 6 authorization from the vendor ~to institute and
- implement elevated pH, coordinated Lithium / Boron chemistry.
(The use of elevated pH chemistry has been shown to minimize radiation field increases in European power plants.) As a result of the licensee's efforts dir, cussed above, San Onofre was well below the 1987 average collective dose for PWRs of 371 person-rem per reactor. Despite having major outages at all units, the average collective dose was 232 person-rem per reactor. This also surpassed the 1990 INPO occupational exposure goal of 288 person-rem per reactor. In addition, the licensee surpassed the 1990 INPO solid radioactive waste goal of 213 cubic meters per reactor by producing only 109 cubic meters per reactor for 1987.
- 2. _ Conclusion Performance Assessment - Category 1.
h < u _* 4. , f.. .- .* a 13 0' 3.- Board Recorrrnendations i
'The'11_censee is encouraged to continue efforts to expeditiously.
. resolve problems _with process and effluent monitoring instrumentation and with safety-affecting-software validation,
~~ verification, and documentation; and to assure active participation Physics progr. am -of all site organizations in a quality Health C. Maintenance / Surveillance
- 1. ' Analysis During the.SALP period, approximately.1260 direct inspection hours-were~ap Surveillance. plied in the area of Plant Maintenance and
- Strengths were observed in'the scheduling and performance of surveillance, implementation of the chemistry based maintenance system.. program, and the effective use of a c Weaknesses identified during the
. period primarily involved procedural deficiencies (i.e., lack of.
detailed work instructions and acceptance criteria).and procedure compliance by maintenance personnel.-- < The NRC routinely monitored licensee maintenance and surveillance activities, ' paying particular attention to the adequacy of issued procedures and compliance with those procedures. Evaluations were also made of the adequacy of licensee programs to ensure followup and trending of failed surveillance, proper clearance of equipment timely perfomance of required. maintenance and surveillance, pr,oper quality.
~ control of safety related materials, and adequate ~
post-maintenance testing.- % specific strength was noted,ip the scheduling of surveillance: N that very few were missed of-several thousand required u '- perfomed during the period. Staffing'of maintenance and sur<etilance activities was considered adequate. The SCE staff exhibited superior performance in water chemistry - control during this assessment period. The licensee was effective in identifying and reducing impurities in secondary water systems, such as in~ limiting dissolved oxygen ingress for protection of condensate and feedwater components. The licensee was also considered an industry leader in the use of in-line ion chromatography methods for continuous measurement of secondary water ionic impurities at the ppb level. Licensee management was actively involved in the scheduling and ' coordination of maintenance and surveillance activities, and the licensee was considered to be responsive in addressing NRC concerns. Significant industry leadership was shown in initiatives related to preventive and predictive maintenance activities. Action was also taken to in: prove reactor coolant system'(RCS) isolation valve leak rate surveillance procedures, improvements were made in station rigging practices, procedural
E ., l i f. . 14 l
. l l
changes were made to improve surveillance of penetrations during mid-loon operation, and several improvements were made in hydrostatic testing practices. In addition, the licensee took timely action to resolve concerns expressed in the previous SALP report relative to control of accelerated maintenance activities and trending of surveillance activities. However, with regard ; to the latter, considerable involvement was required by the ' licensee's QA organization before an acceptable program was developed by the station. A principal weakness observed during this SALP period involved procedure compliance by maintenance personnel. Inspection activities identified four violations involving failure to follow procedures. One Severity Level IV violation applicable to Unit 2 was cited for failure to comply with maintenance procedures for control of measuring and test equipment. Two Severity Level IV violations applicable to Unit 3 involved
- failure to comply with maintenance procedures for transfer of water to the spent fuel pool and failure to comply with an engineering surveillance procedure during containment integrated leak rate testing. In addition, a Severity Level IV violation applicable to Units 2 and 3 involved failure to comply with procedures for documenting nonconforming conditions during the conduct of maintenance activities.
~
Weakness was observed at times in the control of maintenance
' activities. One notable example involved maintenance work inside the Unit 3 pressurizer, which required the reactor coolant system to be drained to mid-loop. Without questioning the possible effect, maintenance personnel working inside the pressurizer inserted a mounting device for a videocamera (used for radiation exposure control) into a pressurizer nozzle.
Since the reference leg tubing for the reactor vessel lete1 indicating system was connected to this nozzle, this caured the reactor level to be indicated incorrectly as the level was being lowered. A potentially serious problem was averted by the alertness of the control room operators, however, as discussed in Section IV.A. Plant. Operations. The NRC also noted a number of additional examples of inadequate procedures and inattentiveness on the part of maintenance personnel. For example, a Unit 1 emergency diesel generator was inadvertently started as a result of inattention to equipment clearance bound ries; numerous foreign material exclusion (FME) problems were encountered during the Unit 3 refueling outage; steam generator cold leg channel heads tere overflowed on Unit 3 when maintenance instructions were not adhered to; and welding rods were not properly controlled during pressurizer heater replacement work on Unit 3. Improvements were noted in housekeeping during maintenance activities, but additional improvements are warranted during major outages. l
]
l l l E__ _ _
.,%; m. . z ' ...
.15
.During the SALF period, there were a total of 21' LERs issued in
.the area ~ of maintenance and surveillance. Of these 21 LERs, 11 involved personnel error and 8 involved inadequate procedures.
Oi.ly 3 of the LERs' involved procedure noncompliance. The LERs adequately described the major aspects of the events and the corrective actions taken or planned to prevent. recurrence.
' 2. . Conclusion Perfonnance Assessment - Category 2.
L
- 3. Board Recommendations Licensee management should continue to emphasize a high. standard of. performance by maintenance supervision'and maintenance personnel.
Measures for exercising control over the conduct of' maintenance activities should be strengthened.. The licensee should also continue efforts to improve the quality of maintenance and surveillance procedures and to ensure complete-adherence to them. Site management should focus special attention on documentation and evaluation of discrepe.nt conditions, initiation. and on the criteria used for nonconformance report D. Emergency Preparedness
- 1. Analysis Region V conducted two emergency preparedness (EP) inspections during this appraisal period. One inspection addrested followup on previous inspection findings and the other addressed'the.
routine inspection program. An annual emergency exercise.was not observed during this SALP period. Approximately 60 Cobrs of direct inspection effort here expended in the EP functional area. Strengths identified during this assessment period were j
. management support of the EP program, organization ard staffing levels of EP personnel, and use of industry events te make program enhancements. o One waakness was identified with re to the effectiveness of training in the EP functional area.gards The inspections conducted during this appraisal period showed a ;
significant strength in licensee management supoort of the EP program. ! Resources have been used to upSrade the Interagency Telephone System, to provide a card reader t.ystem for the Emergency Operations Facility (EOF), to improve accountability, and to redesign the Technical Support Centers to improve information flow. ' A strength was also identified in that the licensee has demcnstrhted initiative in the handling of technical issues, particularly when operational events have occurred. For example, the licensee revised the emergency classification ) procedures to include emergency action levels (EALs) which addreas the loss of Reactor Coolant System (RCS) heat removal [ L G_ \
9 _. ca m g7 .m 16-y
" capability,; and ~ to address situations ~wherein: the plant . H L conditions meet the criteria of an EAL,'but the operational mode does not apply..--The fact that the licensee revised these-j i
procedures as a result of two events (one' occurred at. SONGS'and- Ll the other occurred at another Region V facility):showed that the. j' licensee. recognized the benefits' associated with lessons ~1 earned' ' from industry and their application to San Onofre. Another strength was identified in that SCE has shown '
. improvement in responsiveness to.NRC initiatives. During the
'licenseu's 1987 annual EP exercise' problems associate'd with. d
~
exercise control and over-simulation were~ identified. .Since- 1 then, the licensee developed'a formal-drill' controller trainingc -l program and adopted methods (i.e., the use of props) to increase ? realism.during drills and exercises. . Weaknesses identified:
-during the11987,' exercise involved contamination control in-the Operations.Sunport Center, notifications 'of. in-plant workers,
- and radiological controls in the EOF. . Results from tiic.1988 -
exercise,.which was conducted.in October just outside the SALP1
-period, indicated that the licensee's-cor,rective actions taken i after the-'1987 exercise were effective. 2
.1 A weakness' involving EP training was identified during this ',
assessment period. Inspections conducted during this- appraisal period indicated that the licensee's training program for emergen::y response personnel needed critical examination. The j
- licensee had a training program that included computer based '
. instruction (CBI). This training was coupled with a' quarterly integrated drill' program to provide experience in handling EP q related events. However, despite these programs, interviews with a number of Shift Superintendents revealed weaknesses.in ]
l i their knowledge level and licensee performance during the 1987
- exercise showed a slight declining trend. In response t'o'this-weakness..recent discussions between licensee training' personnel and the NRC revealed that' the CBI portion of the training-l 'l , program was being revised to be more performance based.- It was 1
- - considered that'this effort and the action.taken to increase !
realism during drills should improve the quality of training in j t the.EP area. i H One violation primarily associated with the Safety Assessment /. I Quality Verification area was also related to Emergency Preparedness. This violation, identified during an Emergency
- L Preparedness inspection, involved the failure of the Quality L
A',surance organization to perform a required 12-month audit of Emergency Preparedness. This indicated a'need for additional QA Ll ' comitment to the EP program. i P' i E During the appraisal period, some staffing and organizational p changes occurred that affected the EP Division. In particular, the station EP organization was changed to functionally report 6 to the Operations Department and a new manager was assigned to the Nuclear Affairs and Emergency Planning (Corporate EP) y organization. It is considered that both of these changes have L f 4 -x _ _
f '. t,.p c.~ M, 17.
- m. ,
i 3 i had positive effects. Corporate and Station EP. have been. working stable. well as a team and the staffing has appeared to be'quite
- 2. Conclusion Performance Assessment Category 1.
o
- 3. Board Recommendations-
^
Licensee management is encouraged to continue improvements.to ~ the EP training program. In addition, licensee management is. encouraged to maintain a consistent association between the.EP and QA ' organizations as a result of the failure to audit EP activities. , , E., Security 1
- 1. Analysis
'During this SALP assessment period. Region V conducted three physical security inspections at the San Onofre Nuclear Generating Station.. A total of approximately 240 hours of
< direct-inspection effort were expended by regional. inspectors. In addition, the resident inspectors provided continuing observations in this area.. There were no material control and-accounting inspections conducted during this assessment period. Significant strengths' identified included management involvement, in' activities .that led to the reduction of security events the experience levels and effectiveness of the licensee's , and security staff. The previously identified Regulatory-Effectiveness Review'(RER) finding pertaining to specifig yital area barriers remains unresolved pending a change in NRC requirements. u; A strength evident during this assessment period was the licensee's ability to the overall security program at San Onofre. In maintain addition, athe high assuranc involvement this of the quality was licensee's Station management in' assuring evident. ; The resources available to manage and maintain this program were fully adequate and effectively { j utilized, and resulted in an overall high level of perfomance. The procedures for the Security Division were complete, well stated and explicit. The licensee's remedial measures to correct self-identified deficiencies were effectively addressed 1 in the root cause assessment for each deficiency, and actions have provided lasting corrective measures. Of particular note was the licensee's expansion and improvement of their established Centralized Screening Program. Background screening was completed for all contract employees (as well as licensee employees) seeking access to the protected area. This expanded background screening included even those contract employees who arrived on site with an employment verification letter. As a result, the licensee's expanded efforts exceeded the minimum requirements of the approved security plan and improved the overall quality of the security program. . i
kg.g "*b"'
.18- > ,'
~
~ !
1 . 1 rAnother' strength. identified during this~~ period was security l, management's " continuing. efforts ^ to effectively. coordinate with - other plant staff in- the identification and resolution!of l 1 safety / security c'o ncerns at San Onofre. On-duty plant operators continued to carry an accountable set of keys.for all' locked and :i " alarmed vital areas, which' ensured their:immediate entry to' all! vital areas in the event of an emergency. The experience and. effectiveness of the licensee's security) staff: supporting the'overall security program was considered a { strength. Key positions were identified land responsib111 ties.- were well defined. The Security Department's. Training andl ; Qualification program was effective
- implemented with dedicated resources,. well defined. and During annua 1Lrefresher training, a high degree of realism was achieved'through use of..
,. MILES '(Multiple Integrated Laser Enhancement: System); >r laser-equipped weapons. No viola'tions ay inst the security program were cited during .. this SALP period,'and the licensee reported only eight security events. These numbers showed a significant reduction in comparison to the previous SALP period in which'three. violations. were identified and"115 security eveats were reported. The-eight security events. occurred after a change in the requirements of 10 CFR 73.71(c). : As a. result, they were
-e reported'in the Licensee Event Report (LER) format. These c ' events were security computer failures-(3),' drug-related events (2), loss of security keys '(1), unlocked vital- area: portal'(1),
andmiscellaneousevents(1). . The licensee's a measures, based upon their root cause analyses,pplied appeared correct'va
. complete and effective.' '
In September 1984, prior to the August, 1986 NRC policy
- statement on Fitness for-Duty of nuclear power plant personnel, the licensee implemented a Substance Abuse Program. As
. initially screening implemented, this program included random drug tests. However, in January 1987,' a Federal District Court issued an injunction which limited the licensee to conduct -
only announced annual drug screening tests. With this injunction still in effect, the licensee's Drug Screening Program at San Onofre consisted primarily'of Pre-Access Drug Screening, Annual Drug Screening, For-Cause Drug Testing and an Employee Assistance Program. Additionally, the licensee has. expanded this Program to include the use of drug detection dogs inside the protected area, and random searches of employees and their vehicles when entering the owner controlled area. During this assessment, four information notices related to security were issued. The licensee's actions in response to these notices, were found to be appropriate.
- 2. Conclusion Performance Assessment - Category 1.
,___m ___m------ --- - " - - ' - - " "
. .. l f 19 i 3
1
- 3. Board Reconrnendation
)
Licensee management is encouraged to continue their effective support of the overall security program. J F. Engineering / Technical Support
- 1. Analysis .
During the SALP period, approximately 580 hours of direct inspection Support effort were applied to the Engineering / Technical. area. In addition to continuing coverage by the resident inspectors, a' regional Safety System Functional Inspection (SSFI) team perfonned an inspection in this area. The major weakness in this area involved the discovery of significant inadequacies in the control of design and engineering work, largely resulting from a poorly defined plant design basis and insufficient attention to plant design details. In contrast, a strength observed during the latter part of the SALP period
~
involved the self-critical attitude demonstrated be senior SCE management in acknowledging the need for improved performance in this area, and the planned engineering reorganization, which has been initiated to provide the needed improvements. . The SALP Board considered the licensee to have a capable corporate engineering staff. Improvement was perceived in the quality of engineering work performed during the latter part of this assessment period through the self-imposed evaluation of several safety systems. Increased licensee and NRC emphasis on the quality of engineering activities led to the identification of notable weaknesses which were manifested in several significant' safety-related engineering problems. . . Specific examples included several single-failure vulnerabilities in Unit 1 ESF systems; excessive post-accident loading (in excess of Technical Specification Ifmits) of' Unit I diesel
, genera tors; excessive loading of Unit I charging (due to incorrect use of pump perfonnance curves); pump motors inadequate 18-month testing of safet Nuclear Safety Concern);and y related batteries (in the programmatic response breakdown of to a design controls associated with environmental qualification (EQ) of Unit I electrical equipment (resulted in a $150,000 civil penalty).
The principal causes of these various problems were inadequate administrative controls governing engineering activities, insufficient attention to the quality of engineering work, inadegaate documentation and understanding of the plant design basis by cognizant engineering and technical personnel, and limited engineering resources. Although station and corporate management were involved in engineering work and in the resolution of engineering problems, they were not fully effective in the overali implementation and coordination of engineering and technical work. The SSFI conducted by the NRC in May - June 1988 identified further weaknesses in the licensee's controls affecting _ - - - - ._ a
e..y. - p ;,
.20 E
engineering and technical work. . The results of this inspection, which ~ assessed the operational readiness of the component cooling water (CCW) and salt water cooling (SWC) systems, indicated that SCE did not' fully understand the basic design of the systems reviewed; did not have ready access to accurate system design infonnation; ~and had not performed engineering' work in a complete and technically accurate manner.- The licensee was generally' responsive to NRC initiatives. ' An. example noted during the period was the engineering' evaluation of several important plant systems which SCE performed in advance of the SSFI. This comprehensive evaluation identified many of the deficiencies subsequently noted by the NRC's inspection. In addition to the engineering problems discussed above, the S3FI team and other inspections observed weakness in the interface between the Operations and Enginet ring / Technical organizations which resulted in extended periods needed to
- resolve plant system problems. Examples in;1uded problems with the Unit 2/3 CCW system, low Freon levels in Unit 2/3 emergency chillers, and repetitive and generally spurious actuations of ESF systems and cable spreading room deluge systems. The SSFI team also concluded that the licensee had not reported, as ',
required,'three different deficient conditions associated with the CCW and SWC systems. E While the staffing devoted to the Station Technical organization appeared to be adequate, the SSFI findings and other observations indicated that the' corporate organizations relied heavily on contractors for the accomplishment of engineering q work, particularly on Units 2 and 3. This resulted in some. cases in a loss of corporate memory on system dr in considerations due to turnover of cognizant con; tor 4 personnel. Accountability for engineering ' work .s also ]q
- lacking, with corporate engineering assets reporting to three j different vice presidents. Whfie effective technical training i was provided in some areas, it was'noted to be deficient in !
others; e.g., the SSFI team noted that engineers had ; insufficient knowledge of how and where to obtain available design information. ! NRC inspection efforts identified six enforcement items related to the Engineering and Technical Support area. These included a i Severity Category B EQ violation ($150,000 civil penalty), as I discussed earlier; two Severity Level IV violations involving design and testing deficiencies in the Unit 2/3 CCW and SWC systems; one Severity Level IV violation involving improper 4 separation of electrical cables; one Severity Level IV violation associated with improper testing of Unit 2/3 main steam safety valves; and one Deviation involving improper installation of Unit 2/3 CCW system radiation monitors. e
,m____ __ -- - - - - - -
m
- 0. u[ f, * '
21 + l, O p q C q ;9 ., s A total of.31 LERs were associated with Engineering and - < Technical Support-activities. Morethanhalfofthese-(18)^ involved spurious actuations of engineered. safety' features (ESF),: including containment, fuel building : toxic gas,'and-control room isolation systems. The. remaining 13 LERs involved: violations of plant. technical-specifications or degraded plant safety 'resulting from system design inadequacies'or errors by engineering and . technical . support personnel. . In response to the SSFI findings and the-significant problems: discussed above. SCE management undertook a major reassessment- j 3 of the engineering and technical organizations and the controlst # and methods used in their accomplishment of engineering work. This led to several significant. recommendations which were being L implemented as the SALP period closed. These included (1)"the
-. consolidation of all' corporate engineering. assets under. a. single -
vice president; (2) relocation'of the engineerin D to Irvine, significantly closer to the' site;-(3)g strengthening organization , of in-house engineering capabilities to permit-less' reliance on
. contractors for engineering / design work; and (4) a comprehensive review and updating of.the plant s design basis documents. The-
> licensee expects these actions'to significantly improve the quality of engineering and technical work.done by SCE.:
- 2. Conclusion-Performance assessment - Category 3.
- 3. Board Recommendations The licensee is encouraged.to expeditiously complete the
~
implementation.of identified improvements'in'the corporate engineering organization. and to ensure that necessary and' accompanying improvements are made-to administrative controls affecting engineering.and technical work. Plans:for updating 4' the plant's design data-base and strengthening in-house-engineering capabilities should also be aggressively pursued. G. Safety Assessment / Quality Verification
- 1. ~ Analysis During the SALP period, approximately 860 hours of Crect.
inspection effort were applied to Safety Assessment / Quality Verification. Some strengths were noted during the SALP period, , i predominantly in improvement of the root cause evaluation process and in the initiation of proactive measures to monitor and improve plant performance. However, several significant 1 areas of weakness were noted in this functional area, including I insufficient QA involvement in identifying significant problems, inadequate safety reviews, improper deportability determinations, and inadequately supported amendment requests. b
.o.' '
3' n.}gg ,
'22 y . c
. Several significant weaknesses associated with_ licensing F activities were noted during this period. These i_ndicated insufficient understanding of NRC requirements or the plants' licensing basis, or a lack of thoroughness in the preparation'of licensing submittals, or a non-conservative approach to the-resolution of safety issues. Examples . included the followingi.
N The licensee did not demonstrate a thorough' understanding of how to apply the regulatory requirements specified in 10: CFR 50.59 to the licensing' basis of-the. units (e.g., the licensee's inappropriate handling of the proposed 4 transshipment of spent fuel from Unit I to Units 2 and 3). SCE's submittals to NRR were frequently late. . Examples of late submittals: included responses to requests for
' additional information concerning the spent fuel transshipment,- the proposed nuclear instrumentation upgrades, the Unit I cask drop analysis, ESF single failure information, information concerning TMI item III.D.3.4, and-five items concerning the Systematic Evaluation Program
.(SEP).
The' licensee notified the NRC in September 19T6 that a ' report of reactor vessel specimen test results would be
- 1 ate. The specimen was removed on September 20, 1987, but.
the letter was not sent to the NRC until September 20, 1988. The extension required by Appendix H to 10 CFR 50 was not requested. The licensee was slow to respond to NRR reconrnendations j that a " slow" start (24 seconds or longer) be used for all Sl Unit 1 TDI diesel generator. starts performed for . i R maintenance or surveillance purposes. The purpose bf the recommendation was to minimize transient stresses on'the crankshaft, which was vulnerable to cracking at the ;
. lubricating oil holes. NRR subsequently required i
crankshaft inspections to be conducted, and made slow starts a license condition-in August 1988. j
- .3 In response to main steam isolation valve (MSIV) failure at another facility which demonstrated a possible common mode failure mechanism, SCE performed a boroscopic examination of a Unit 3 MSIV and a root cause analysis of the i) failures. However, SCE was reluctant to disassemble a Unit 3 MSIV even though Unit 2 (in' power operation) was {
a also potentially affe.;ed. After SCE was persuaded to disassemble one of the MSIVs, the findings did not support 1 i the results of their boroscopic examination. Consequently, l the initial reports of these two efforts were contradictory. i l!
+j,
. 23 ']
n In 1981, Unit I experienced a common-mode failure of the hydraulically-operated safety injection pump discharge valves, and subsequently committed to study long-term design improvements. This comitment was subsequently withdrawn, however, based upon a cost-benefit analysis, and SCE did not propose a cost-effective alternative until encouraged to do so. . Inspection activities during the period resulted in the' l identification of six enforcement items. Specific enforcement i topics included one Severity Level IV violation for failure to I maintain a feedwater isolation valve operable; one Severity Level IV and one Severity Level V violation for failure to j perform required quality assurance audits.(involving the .{ emergency preparedness and radiation protection areas); two Severity Level IV violations (one with 3 instances) for failure to make required licensee event reports; and one Deviation for siphon to failure implement an FSAR commitment for spent fuel pool breakers. The violations involving failure to make required reports indicated that excessive attention was given to establishing that a situation was not reportable rather than i conservatively reporting it and supplementing or canceling the report when analyses were completed. Some enforcement actions discussed under the Engineering / Technical Support area also reflected on this area due to insufficient or untimel
- involvement by QA and/or licensing personnel -- e.g., ythe Unit 1 j environmental qualification violation and the Unit 2/3 CCW system design violations.
A total of 12 LERs were associated with Safety Assessment / Quality Verification activities. All but one of these LERs were primarily applicable to the Engineering and Technical Support functional area. However, they also reflected adversely on this functional area, since they involved missed opportunities for the licensee's quality assurance organization and safety oversight groups to identify and correct the problems. These events included: Unit I single fail'ure problems s Unit 2/3 CCW design problems Unit 2/3 steam safety valve setpoint problems Unit I diesel generator electrical load problems Unit 2/3 battery service test problems Unit 2 spent fuel pool siphon problem Unit 2/3 emergency chiller Freon problems Unit 1 environmental qualification problems The NRC observed some positive initiatives by SCE during the SALP period. For example, the licensee undertook an ambitious effort to monitor the performance of safety-related instrumentation, with the ultimate goal of establishing a j reliability-based surveillance requirement. The program i appeared to be well thought-out and should contribute to l industry /NRC efforts to improve Technical Specifications. The I l I< l
. .w ' 3 p 4..;' .y-y licensee was also cooperative with the_NRC in'an information exchange related to'an NRC study on Technical Specifications surveillance requirements. Another licensee initiative was the establishment of a performance-based inspection training -
program for QC inspectors, similar to the methodology used by. the NRC to increase inspection effectiveness. 'For Units 2 and' 3, the licensee initiated a program which usesLa-generic
.probabilistic risk assessment (PRA)-study to detemine the safety gains to be realized from improved system reliability.
Quality ' program activities appeared to be adequately stiffed, and the' licensee made progress in correcting deficiencies ' observed during the previous SALP period. For example, the licensee's root cause assessment program was overhauled and. appeared to be more effective. in identifying and correcting. the root causes of plant events. Also, the licensee implemented an
- -( -
extensive audit of the design control process which identified- 1 several significant problems and recommended organizational and . other changes to provide improved performance. During the SSFI that was conducted in May and June 1988, the. team observed activities of several of the quality oversight groups in order to determine their effectiveness.. These groups were the Nuclear Safety Croup (NSG), the Nuclear Control Board
- (NCB), the QA organization, the On-Site Review Committee.(OSRC),
1 and the Independent Safety Engineering Group (ISEG). As a result of this review, the team found-that the NSG and OSRC were conducting adequate reviews of plant activities, so that technical specification requirements were being met. The
- NCB (not required by Technical Specifications) complemented NSG activities by providing a' vehicle for senior management oversight of nuclear safety functions. The site QA group had recently initiated a plan to conduct detailed technical audits, which initially included an extensive design control audit involving three full-time and eight part-time auditors for more than 5500 man-hours. This audit identified 71 needed corrective actions. The SSFI found that the ISEG was effective in fulfilling its functions as described in the technical specifications and had exercised some proactive influence for the betterment of plant operation and safety by early identification of problems.
Significant problems that were identified during the SALP period indicated the need fem closer evaluation of oversight group perfomance. In that regard, shortly after the end of the SALP period, the NRC perforined a review of QA audits and surveillance that were conducted in 1988. The review indicated
' that some significant problems were identified during the perfomance of these audits and surveillance. However, for the most part, findings were of minfinal significance and there was a perception that QA was not sufficiently aggressive in probing to the depths necessary to effectively assess the adequacy of programs.
. _______.t_.m.----------" '
Y , . ? L. . gg l b As noted above, principal shortcomings in this area during the b SALP period were weaknesses in licensing activities and insufficient involvement by the quality assurance organization }; and safety oversight groups in.the plant engineering area. 1 Almost' every. inquiry into this area by the NRC or the licensee identified significant weaknesses in the control and implementation of engineering work. The Board acknowledged that senior licensee management had recognized this deficiency and-implemented actions to correct the basic problems. The Board noted that the recent restructuring of the licensee's Nuclear i Engineering, Safety, and Licensing (NES&L)-Department also { changed the organization and management of the various quality: '{ assurance and quality oversight groups. The potential gains , resulting the from period. next SALP these changes will be evaluated closely during
- 2. . Conclusion {
I l Performance assessment - Category 3.
- 3. Board Recommendations The licensee should give significant additional emphasis to insightful definition and a audits and safety reviews. ggressive performance of quality Management should focus attention on effective implementation of the NES&L reorganization and ,
other actions to improve the weaknesses discussed. More thorough review should also be provided for. licensing submittals to ensurebases. design proper consideration of NRC requirements and applicable V. SUPPORTING DATA AND SUMMARIES A. Enforcement Activity
- Three period.
resident inspectors were onsite during the SALP assessment Thirty-six inspections, including a team Safety System Functional Inspection (SSFI) in May and June 1988, were conducted during this period for a total of 5437 inspector hours. A sunnary of inspection activities is provided in Table 1 along with a summary of enforcement items from these inspections. A description of the enforcement items is provided in Table 2. During this SALP period, one escalated enforcement item ($150,000 civil penalty) was identified, concerning environmental qualification of Unit 1 safety , related electrical equipment. ] j B. Confirmation of Action Letters l No Confirmation of Action Letters were issued during this assessment period. 1
)
1 i 1
i.;-> l y ,;l ' ' .' - 26 ,
- j. ':
e . C. 'Other
' An Office for Analysis ,and Evalu'ation of Operational Data (AEOD) review of licensee events at. San Onofre is included as Attachment 1..-
AEOD reviewed the LERs and significant operating events for quality of reporting and effectiveness of identified corrective actions. S ' 9 e O 6 e e
- F e
l ( l l _ _ _ = _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - - - _ _ - - - -J
L. ,. . . / P , , l.- "; l- _ TABLE 1-p 1NSPECTION ACTIVITIES AND ENFORCEMENT
SUMMARY
Functional Inspection Enforcement Items *
- Area Percent Severity Level-Hours _of Effort. I II III IV V- D***
A. Plant Operations 1802 33.14 l' B. Radiological 622 11.44 Controls '2 1 C. Maintenance /- 1262 23.21 4 Surveillance D. - Emergency Prep. 60 1.10 E. . Security 247 4.54 F. Engineering / $84 10.74 1** 4 Technical Support 1 G. Safety Assessment / 860 15.82 Quality Verif. 4- 1 1 Totals 5437 100.00 1 15 2 2 Severity levels are discussed in 10 CFR 2 Appendix C. Two deviations (one each in areas F and G) were identified during this SALP period. This violation was a Category B violation concerning EQ. ** Denotes deviations discussed 11 Table 2 This information is current through inspection reports 206/88-23; 361/88-24; and 362/88-26. I-O __m____ _ . _ _ _ _ _ .- - - - - - - - - - - - ' - - ^ ^ - - - ' - - --
. ,. ji
- j
' TABLE 2 f
ENFORCEMENT AC IVITY' Unit 1 Inspection-Report No. Severity Functional. Subject. Level
, Area -
88-03 Failure to make proper safety system IV A' operability determinations
- 88-06 -
Failure to post a high radiation area IV 8 88-07 Failure to conduct an audit of' the # .V G Emergency Preparedness program 88-10 Environmental qualification deficiencies 8 F/G 88-23 Whole body exposure in excess of the-IV B quarterly limit
#' Applies to Units 1, 2, and 3.
, Unit 2 Inspection Severity Report No. Subject functional Level Area 87-25 Failure to' post a radiation area V 8 87-31 Failure to report steam generator ## IV *E
_ safety valve inoperability. 88-03 Failure to document nonconforming ## IV C conditions during maintenance 88-03 Failure to comply with' Technical ## IV F Specification requirement for' testing main steam safety valves 88-10 Failure to report component cooling ## IV G water system design-deficiencies B8-10 Failure to' include analyses of ## IV F/G adverse effects of earthquakes on the design of equipment _ _ , . .2., _ _m._------ - - - - - - - - - - - "-'
g.S'*'. 1 4 Table-2, Enforcement Items (Continued) Inspection Report No. Subject Severity Functional level Area 88-10' Failure to include saltwater cooling ## IV. F valves in the in-service testing program . 88-10 Deviation - Mode of operation of ## F component cooling water provides no ' monitoring ability for the loop containing the letdown heat exchanger 88-15 Inadequate control of M&TE (two examples) IV C/G 88-15 Deviation - Fuel pool purification piping ## G not installed in accordance with the FSAR
.. 88-18 Train A and B cables in direct contact with IV F one another in a post accident panel
## Applies to Units 2 and 3.
- Unit 3 Inspection Report No. Subject Severity Functional Level Area 87-25 Continued operation with a main feedwater IV *G' isolation valve and ADS valves inoperable 88-04 Inadequate QA audit program for radioactive IV G transportation packages 88-20 Failure to comply with ' procedures for IV C temporary spent fuel pit transfer pumps 88-22 Failure to adequately control the IV C performance of an integrated leak rate test ,
Functional Areas A - Plant Operations 8 - Radiological Controls C - Maintenance / Surveillance D - Emergency Drep. E - Security F - Engineering / Technical Support G - Safety Assessment / Quality Verification
' <J g to; .:
f 'c (l'
- TABLE 3A -- L' nit 1 SYNOPSIS OF LICENSEE EVENT REPORTS (LERs)
Functional SALP Cause Code
- 1 Area A B C D E X Totals - - - - - -
- i. A. Plant Operations 1 2 ".3 B. Radiological 1
~
Controls 1 C. Maintenance / 4 2 1 7 Surveillance
; D. Emergency Prep.
r E. Security 1 1 .
. F. Engineering / 4 1 Technical Support 5 G. Safety Assessment / 1 Quality Verif. 1 Totals 9 1 3 4 1 18
- Cause Code A - Personnel Error ,
B - Design Manufacturing or Installation Error C - External Cause ** D - Defective Procedures E . Component Failure X - Other Functional Areas A - Plant Operations ' B - Radiological Controls C - Maintenance / Surveillance C - Emergency Prep. E - Security F - Engineering / Technical Support G - Safety Assessment / Quality Verif. The above data are based upon LERs 87-15 through 88-14
. g .. ,,j
- e :
'*[
-TABLE 3B - Unit ?
SYNDPSIS OF L ICENSEE EVENT REPORTS'(LERs)' Functional Area SALP Cause Code *
~A B C D
- Totalf ~ ~ ~ ~
E
~~ ~
X A. Plant Operations 2 3 g B.- Radiological -l 1 3 Controls 2- 'i i C. Maintenance / :\
$ 5 Surveillance '30 ,;
D.- Emergency Prep. E. Security-F. !
.. Engineering / 5 18 o Technical Support 23 G. Safety Assessment /
Quality Verif. Totals 13 18 6 3 40
- Cause Code A'- Personnel Error B - Design, Manufacturing or Installation Error C - External Cause **
D - Defective Procedures E - Component Failure X - Other Functional Areas < A - Plant Operations B - Radiological Controls C - Maintenance / Surveillance D - Emergency Prep. E - Security F Engineering / Technical Support G - Safety Assessment / Quality Verif. The above data are based upon LERs 87-22 through 88-26. l-u___ =_ _
1 l t. s . '* ' : . . l /: 'S-
- j. ,
i TABLE 3C - Unit 3 i SYN 0PSIS OF LICENSEE EVENT REPORTS (LERs) Functional SALP Cause Code
- Area -A B C D E X
._& Totals- ~ ~ ~ ~ ~ ~
A. Plant Operations 1 .1 2
'B. Radiological 1 Controls *1 C. Maintenance / 2> 1 1 4 Surveillance
/ D. Emergency Prep.
- l" E. Security J. F. Engineering / 1 2 Technicol Support 3 G. ' Safety Assessment /
Quality 1/erif.
. Totals 5 2 1 1 1 10
- Cause Code A - Personnel Error B - Design, Manufacturing or Installation Error C - External Cause *'
D Defective Precedures E Component Failure X - Other Functional Areas A. - Plant Operations B - Radiological Controls C - Maintenance / Surveillance D - Emergency Prep. E - Security F'- Engineering / Technical Support G - Safety Assessrrent/ Quality Verif. The above data are based upon LERs 87-17 through 88-09.
.G'*.
1
,a ATTACHMENT 1 Unit 1 Licensee Event Reports (LERs)
The Analysis Branch of the Office for Analysis and Evaluation of. Operational Data (AEOD) reviewed 17 LERs issued by Southern California Edison, not including revisions, for Unit I during the assessment period from October 1, 1987 through September 30, 1988. The review included LERs numbered as follows: 87-015 to 88-013 The LER review followed the general instructions and procedures of NUREG-1022. The specific review criteria and the findings were as follows:
- 1. Significant Operatino Events ..
The following four or.currences were determined to be potentially. significant by the.AEOD screening process: LER.87-15, concerning single failures of engineered safety features systems pertaining to decay heat removal, main steam
.line break mitigation, and steam generator overfill.
LER 87-16, involving failure of four air operated valves to function due to solenoid valve failures, rendering independent
, trains in multiple systems inoperable.
LER 88-01, referring to environmental qualification prog ~ra'm deficiencies. l
- LER 88-09, regarding electrically loading both emergency diesel generators in excess of the Technical Specification maximum allowable kilowatt loading.
- 2. Causes i
Root causes associated with the 17 events included: Three personnel errors Four procedural / administrative errors Four design / installation / fabrication Six undetermined These events evaluated did not appear to involve related occurrence:, and no causes were found to be prominent. However, on two occasions (LERs 87-17 and 87-18) voluntary entry into Technical Specification 3.0.3 occurred. l l C__ ___ _ _ - - _ _ - - - - - - - - - - - - - - - - - - - - - - - -
C}i 3y. Q R is f ;";,, ' w Attachment.1 (Continued) 3.. LER Ouality-
.e B The. LERs reviewed adequately described 'all'the major aspects of the i events,. including component or system failures that contributed to the event and corrective actions taken or planned to prevent recurrence.
. easy tc understand. .The reports were reasonably complete, well' written and'
. previous similar occurrences were. properly' referenced i However, many LERs(e.g...LERs further investigations indicated the root cause was unknown pefiding
.and88-09). Updated LERs were then to be issued at the conclus
.of the investigations.
As of the date of this. evaluation performed by.AE00, none of the supplemental reports were received by the NRC. f Units 2 and 3 1.- LER Review San Onofre submitted about.34 reports and four. updates for Unit 2 and about promised eighc reports updates for Unit 3 during this assessment period. Unit 2 for LERs 87-02,87-24,88-05,07?08,09,11,13, cand update, 17 88-02. which have not been received.' Unit 3 has one outstanding Our review' included the following LER numbers: Unit.
.. 2, 88-07.- 87-18 to 87-31 and 88-01 to 88-20; Unit 3, 87-17 and 88-01 to
.One the siphoning LER wasofclassified the spentas fuel significant, pool. 88-17 for Unit 2 concerning The causes were the following:
Six personnel errors for Unit 2 and two-for Unit 3
- Four maintenance errors for Unit 2 and none for Unit 3
, - Six design / installation errors for Unit 2 and none for Unit 3 Unit 3procedural / administrative errors for Unit 2 and four for Eight
- Six causes unknown for' Unit 2 and one for Unit 3 Four equipment failures for Unit 2.and one for Unit 3 1
The majority of the LERs were concerned with'actuations of the toxic gas isolation system, fuel handling building isolation system, control room isolation system, and the containment isolation system. ! These problems were recurring and have been for a long time. Because L of this, the arguments for the causes given were not persuasive. That is tonot probably say, the root cause for these spurious problems was known. The LERs adequately described the major aspects of the events, including component or system failures that contributed to the event and the corrective actions taken or planned to prevent recurrence. The reports were well written. Updated LERs provided new information, denoting the portion of the report that was revised by a vertical line in the right hand margin. e
a Attachment 1 (Continued)
- 2. Preliminary Notifications (PNs)
The Region wrote a number of PNs during this period concerning the two plants. been reportable..No LER could be found for three of these.which ray have PNO-V-88-022 Reactor Shutdown Caused by Increased Steam Ginerator for Unit 2 . Tube Leak. PNO-V-8-002 Unit 3 ' Reactor shutdown Commenced for More Than 48 H Due to Alarms on the Main Generator Hydrogen Detraining Unit. PNO-V-88-047 Cavitation of the Shutdown Cooling Pump Occurred for Unit 3 During Drain Down of the Reactor Vessel.
- 3. 10 CFR 50.72 Reports ,
A review of reports made pursuant to 10 CFR 50.72 identified no reporting deficiencies. n e 9
p,. <. - QUESTION 8 What was the NRC's opinion of the assessment? ANSWER The cover letter and sunrnary section of'the SALP Board Report, which are attached to the response to Question 7, contain the NRC's evaluation of the assessment. In general, the NRC staff found the performance of.the licensee to be acceptable and directed toward safe facility operation, although weaknesses were noted in several areas, including control of maintenance activities, compliance with maintenance procedures and instructions, in the performance of certain engineering and technical functions, and in implementing the quality assurance program. The s'mmary section of the SALP Board Report also includes-the ratings for the las* two SALP periods and the criteria upon which those ratings are based.
(4 < g w: ' . x QUESTION 9- Why has not the NRC demanded that the aiodifications be made and
' the deadline be met regarding the modifications required by the .
NRC as an urgency measure 10 years ago as a result of the Three Mile Island accident? ANSWER NRC's' TMI Action Plan, published nearly 10 years ago, identified specific problems at the Three Mile Island Nuclear Power Plant that. contributed to the 1979 accident and required all licensees to evaluate their facilities-in light of the TMI experience and to take appropriate corrective action. Even when initially published, the degree of urgency of any single TMI Action Plan Item varied from plant to plant depending on the specific design of the facility,
~the actual operating procedures used at each plant, and the degree to which the item may have already been addressed by individual licensees. Since the Plan was published, some of the TMI Action Plan Items have become less urgent than they initially appeared in the immediate aftermath of the accident, while new problems, of equal or greater safety significance than those identified in the Plan, have emerged.
Consequently, the NRC evaluates the priority of each TMI Action Plan item in conjunction with all other NRC requirements, unique plant characteristics, interim measures, and the demonstrated licensee performance at each reactor facility. The NRC authorizes a facility to restart or begin operation only after a careful review of the facility for compliance with NRC requirements and an evaluation of the licensee's demonstrated capabilities to safely operate f 1
)
)
QUESTION 9 (Continued) l 1 the facility. Occasionally, the NRC has authorized some licensees to begin , operation of new plants or resume operation of existing plants without full implementation of all TMI Action Plan items. In the case of San Onofre Units
. I. 2, and 3, the NRC allowed operation only after ensuring that the licensee either had taken sufficient compensatory measures or had implemented the safety-significant portions of particular TMI Action Plan items so that public health and safety were assured.
The great majority' of TMI Action Plan items have been impicinented at San Onofre
. Unit 1. The staff met with SCE on May 1, 1989, to discuss the licensee's schedule for implementing the remaining items. If the NRC staff judges that final implementation of TMI Action Plan items is being delayed unreasonably, we are prepared to issue orders requiring implementation of certain items. A recent example is the order issued by the Director of Nuclear Reactor Regulation on May 10, 1989, requiring installation of reactor vessel water.
level instrumentation.
f I 1 l QUESTION 10 Why hasn't Edison been required (under the TMI Action Plan) to , complete the control room design review or the Safety Parameter l Display System for San Onofre 17 ANSWER The Detailed Control Room Design Review (DCRDR) has been completed by the licensee, and the NRC staff has recently been discussing with SCE improvements in the implementation schedule for modifications to correct human engineering deficiencies. SCE installed a Safety Parameter Display System (SPDS) in 1981 but is reviewing four options for improving it in response to later NRC guidance. The licensee has proposed a date of June 1993 for completion of the SPDS upgrade activity. The DCRDR and SPDS are two of the open items that are discussed in the answer to Question 9.
)
.f . o OVESTION 11 San 'Onofre 1 has also not been required to install' proper - instrumentation to detect inadequate reactor core cooling, nor has NRC demanded that Edison complete testing and analysis for several scfety-related valves and pipes associated with cooling. Why? ANSWER The instrumentation to detect adequate core cooling is a' reactor vessel water level instrument. It provides useful information to the operators in situa-tions involving loss of coolant (water) from the reactor coolant system. As discussed in the answer'to Question 9 an NRC order dated May 10, 1989, requires SCE to install a reactor vessel water level instrument by the next. refueling outage (spring 1991). On February 24,1989, the 120 staff issued its Safety Evaluation regarding testing of relief and safety velves. The evaluation contained a few minor open items which will be audited by the NRC staff. Any additional actions required of the licensee will be completed by the next refueling outage. j _ _-_-_ -_ k
x;
.i j
f.; , e. , h l, QUESTION 12 How can it be assumed that this reactor can operate safely with L a damaged thermal' heat shield since Edison has not yet ! installed the proper equipment to facilitate monitoring of adequate reactor core cooling? ANSWER As noted in our response to Question'11, the licensee is required to' install a reactor vessel water level instrument to facilitate monitoring of adequate reactor core cooling by 1991. However, the instrumentation has no connection with operation with the degraded thermal shield. The reactor vessel water level instrument provides useful infor; nation to the operators in situations involving loss of coolant (water) from the reactor coolant system but does not provide the necessary thermal shield monitoring for safe operation. Information on the NRC staff's basis for interim operation is provided in License Amendment No. 127 and the associated Safety Evaluation (attached to Question 3). The necessary operational monitoring of the degraded thermal shield is provided by a neutron noise monitor and a loose parts monitor system. These systems were required to be in service prior to restart. l _w__u_.____._________
e i 6, . QUESTION 13 NRC has allowed Edison to escape deadlines for San Onofre's revisions in its accident prevention and response procedures plan. Why? ANSWER This question appears to relate to the Procedures Generation Package (PGP), which provides methods for preparing plant emergency operating procedures (EOPs). SCf. has already prepared E0Ps for Units 1, 2, and 3 based on NRC-approved Westinghouse (Unit 1) and Combustion Engineering (Units 2 and 3)- Owners Group guidelines. In N9 REG-0737, " Clarification of TMI Action Plan Requirements," the Commission proposed implementation of revised E0Ps by the first refueling outage after January 1, 1982. In " Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability (Generic Letter 82-33)," the Comission recognized the difficulty of implementing generic deadlines and allowed licensees to establish realistic plant-specific schedules through mutual agreement with the assigned j NRC Project Manager. Supplement I also specified submittal of a PGP that describes a licensee's program for upgrading and implementing E0Ps. Southern California Edison Company issued the working set of upgraded E0Ps for San Onofre Unit 1 in October of 1983, which met the original date specified in NUREG-0737. SCE submitted the PGP for San Onofre Unit 1 on April 12, 1985, which met the schedule mutually agreed upon by the licensee and the NRC staff. Both the E0Ps and the PGP are continuing to be revised to incorporate necessary improvements identified by the NRC staff in the review of the PGP. ; l I l l
, ;,' w ;in . '
I' ! i '.- ' sy t ^[ .,It :Y ?I '
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,, p
/
l QUESTION 14 Why hasn't' the NRC required Edison.to complete the facility that is- necessary to effectively implement emergency planning 1 l and evacuation procedures (the Emergency.0perations Center)? ,
.] .
l !- A _NSWER L j The Emergency Operations Facility (EOF) for th'e San Onofre site is corrplete;in: all respects.-
?
l E I
'l
_m_.______ __li.____________._. . _ _ . . .}}