IR 05000361/1988010
| ML20206C077 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/07/1988 |
| From: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Baskin K SOUTHERN CALIFORNIA EDISON CO. |
| References | |
| NUDOCS 8811160087 | |
| Preceding documents: |
|
| Download: ML20206C077 (1) | |
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l 0 7 gg Docket No. 50-361 50-362
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Southern California Edison Company
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P. O. Box 800 l
2244 Walnut Grove Avenue Rosemead, California 91770 Attention: Mr. Kenneth P. Baskin, Vice President l
Nuclear Engineering Safety and Licensing Department
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Thank you for your letters of September 2 and October 3,1988, in response to
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our Notice of Violation, Notice of Deviation, and Inspection Report No.
50-361/88-10 and 50-362/88-10, dated August 3, 1988, informing us of the t
actions you have initiated to correct the items which we brought to your
attention. Your short term corrective actions will be verified during a
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future inspection. Your long term actions with regard to the performance of engineering activities were discussed during our management meeting on
November 2,1988, and will be followed as a part of our inspection process.
Your cooperation with us is appreciated, f
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Sincerely
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criimt a vt by, l
r R. F. Kit:6..,
i D. F. Kirsch, Director
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Oivision of Reactor Safety and Projects bec w/ copy of letters 9/02/88 and 10/3/88 i
docket file State of California A. Johnson G. Cook B. Faulkenberry J. Martin Resident inspector Project Inspector
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Southem Califomia Edison Company C3 SEP G A9: 55 R+====
3244 WALNUT OROVE AVENUE RO S EM E AD. CALIFORNI A 91770 v ct p a cto September 2,1988
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Director, Office of Enforcement U. S. Nuclear Regulatory Commission Attention: Document Control Desk Hashington, D. C.
20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Reply to a Notice of Violation and Notice of Deviation San Onofre Nuclear Generating Station Units 2 and 3 Reference:
Letter from Mr. John B. Hartin (USNRC) to Mr. Kenneth P. Baskin (SCE), dated August 3, 1988 The referenced letter forwarded Notices of Violation and a Notice of Deviation resulting from the Safety System Functional Inspection (SSFI) conducted between May 2 and June 10, 1988, of activities authorized by NRC Licenses NPF-10 and NPF-15.
The SSFI assessed the operational readiness of the Component Cooling Water (CCH) and Salt Water Cooling (SHC) systems under normal and analyzed accident conditions. This inspection is documented in NRC Inspection Report Nos. 50-361/88-10 and 50-362/88-10, included with the referenced letter.
In accordance with 10 CFR 2.201 Enclosure 1 to this letter provides the Southern California Edison (SCE) reply to the Notices of Violation.
SCE's response to the Notice of Deviation is provided by Enclosure 2.
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The referenced letter requested that a written description of SCE's action plan to address the basic issues identified by the SSFI team be provided within 60 days.
SCE's response to this request will be forwarded under separate cover by October 3, 1988.
At the SSFI exit meeting, SCE was encouraged to perform a prompt, thorough reassessment of the entire CCH System because of the number of questions and concerns raised by both the NRC and SCE. The results of the initial phase of this effort were forwarded by SCE's letter dated June 24, 1988, receipt of which was acknowledged by the referenced letter.
SCE's June 24, 1988 letter
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Document Control Desk-2-September 2,1988 indicated that the second phase would be completed by thu end c' ieptembir.
86cause our initial reassessment concluded that no unreviewed c'ety questions were involved, and that the CCH System remains operable onder the identified conditions, and in the interest of providing a thorough and sonalete reassessment, SCE finds it nec* :ary to delay the complettorr rt this effort e
untti November 30, 1988.
If you have any questions regarding SCE's response to the Notices of Vio1* tion and Deviation or require additional information, please t,all me.
Respectfully submitted.
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Enclosures J. B. Martin, Regional Administrator, NRC Region V cc:
F. R. Huey, NRC Senior Resident inspector, San Onofre Units 1, 2 and 3 D. E. Hickman, Project Manager, Project Directorate V NRR f
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ENCLOSURE 1 REPLY TO A NOTICE OF VIOLATION Appendix A to Mr. Martin's letter dated August 3, 1988 states in part:
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10 CFR 50.73 requ' es, in part, that licensees shall report any event or condition that re
'ted in the condition o# the nuclear power plant,
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including its pr.: iple safety barriers, being seriously degraded; that
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resulted in th6 0.
aar power plant being in an unanalyzed condition that
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significantly compre.aised plant safety; that resulted in the nuclere power plant being in a condition that was outside the design basir of the plant: or any event or condition that alone.could have prevented the fulfillment of the safety function of 'systeras needed to mitigatr the consequences of an accident.
Contrary to the above, as of June 10, 1988, the licenseo failed to report the following conditions:
(1) The High Energy Line Break A'.ldent (HELBA) analysis not having been adequately performed for the CCH system during plant licensing.
(2) The combination of CCH leakage and valve closure time which could
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have prevented the CCH system from functioning during a HELBA.
(3) The CCH system had leakage in excess of the design leakage and had no capability for seismically qualif tad makeup to the system prior to 1984, as reported in delinquent Licensee Event Report 88-008.
t This is a severity level IV violation (Supplement I).
RESPONSE REASON FOR THE VIOLATION A.(1) SCE's failure to report that the HELBA analysis had not been i
adequately performed for the CCH system during plant licensi..y results from insufficient programmatic requirements for offsite organizations (Nuclear Engineering Safety and Licensing; E.ngineering and Construction) to assess the implications of their offsite activities from a reportability standpoint.
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Reportability determination is generally the responsibilit
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onsite (Nuclear Generation Site) Station Compliance group.y of the
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Effective procedural mechanisms are well established onsite for the identification and evaluation of potentially reportable
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occurrences. Although offsite personnel recognize their
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obligation to report safety problems, inadequate programmatic mechanisms have been established offsite for the identification and evaluation of potentially reportable occurrences related to offsite activities.
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In December 1987 SCE commenced a limited review of CCH system performance.
This activity was conducted by offsite engineers.
During the course of this review, it became hpparent that the impact of HELBA on the CCW System had not been adequately considered by the original HELBA analysis.
This dis otery was not considered from a reportability standpoint nor idenfitted to Station Compliance as being potentially reportablF because of a lack of well established programmatic requirements for the offsite personnel who were conducting the review to consider reportability.
A.(2) Licenst Event Report 88-008, dated April 29, 1988 reported that
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the CCH System had leakage in excets of design leakage and had no provision for seismically quali.fted make-up to the system prior to 1984. During the development of LER 88-008, a significant amount of information contained in a reevaluation of CCH System operability was evaluated for reportability pursuant to 10 CFR 50.73. This reportability evaluation considered the interim corrective actions being taken, including the reduction of the inservice testing (IST) program allowable stroke time of the CCW non-critical loop (NCL) isolation valves. While the CCW System operability reevaluation was being refined, interim correc;1ve action to reduce the IST allowable stroke time was considered necessary in order to ensure sufficient CCH Surge Tank inventory to preclude solid system operation following a HELBA.
As part of the evaluation SCE researched IST re:ords for Units 2 and 3 and determined that the measured stroke times of the subject valves had remained below the newly prescribed value subsequent to
, the discovery of the leakage problem in 1983.
Since SCE's research resulted in the conclusion that both Units 2 and 3 continued to meet the non-critical loop isolation requirements,
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the reportability requirements pursuant to 10 CFR 50.73 were judged not to be applicable. Although it was realized that l
excessive CCH leakage together with excessive NCL isolation valve i
stroke times could prevent fulfillment of the safety function, it
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i we.s judged that, with satisfactory stroke times, the condition
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"alone" was not reportable pursuant to 10 CFR 50.73(a)(2)(v).
l As acknowledged by the NRC in Inspection Report Nos. 50- 361/88-10 and 50-362/38-10, which forwarded the Notice of Violation, and in NUREG 1022, Licensee Event Report System, engineering judgment is sometimes required in evaluating plant events pursuant tc certain paragraphs of 10 CFR su.73. As such, there will continue to be opportunities in which licensee's make decisions based c judgment which are later challenged and found to differ from the NRC's judgment of the reporting requirerents for a given circumstance.
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A.(3) Licensee Event Report 88-008, datW April 29, 1988, reported that the CCH System had leakage in excess of the design leakage and had no provision for seismically qualified makeup to the system prior to 1984.
Excessive CCH System leakage was recognize <1 in late 1982. However, the personnal who identified that actual CCH System leakage exceeded the system leakage criteria, failed to initiate a N Mconformance Report (NCR) as require ~d by procedure.
The NCR process is the mechanism by which a reportabi y
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evaluation would have been made for such conditions. A Startup Problem Report (SPR) which addressed the issue was initiated.
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existence of an SPR on the subject may have contributed to the l
reason an NCR was not initiated.
t Because an NCR was not initiated for err.essive CCH sptem leakige,
the situation would not have been formally evaluated for
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reportability. Additionally, in the 1982/83 time frame, the NCR process was being enhanced and was in transition with regard to the inclusion of operability /reportability assessments as part of the NCR form. The NCR transition period is evident by the review
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of NCR's written on Reactor Trip Breakers (RTB) in the 1982/83 time frame.
Some RTB-NCR's written during the trial period of
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this enhancement to the NCR program included operability /reportability assessment while others did not.
j CORRECTIVE ACTIONS THAT HAVE BEEN lAKEN AND RESULTS ACHIEVED A.(1) The availability of the NCR process to document design related problems has been emphasized to offsite engineering personnel.
A.(2) Station Compliance personnel taking part in the reportability evaluation process have been provided augmented guidance for such determinations.
Included in this guidance is reference to the use of engineering judgment and examples of the use of such judgment in reporting determinations.
A.(3)
Since 1983, the NCR process has matured, reducing ths potential for conditions, such as those reported by the delinquent LER, not
being properly assessed for reportability.
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l CORRECTIVE ACTIONS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS
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A.(1) SCE will expand the use of the NCR process by offsite organizations.
Training on this process will be administered to appropriate offsite supervisory personnel, and will in:lude guidance necessary to adequately evaluate findings to ensure that NRC reporting requirements are properly considered.
This training
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will be implemented by October 30, 1988 and it is anticipated that the training will be completed by November 18, 1988. Additional future corrective actions will be discussed in SCE's action plan to address the basic issues identified by the SSFI team, which is to be submitted by October 3, 1988.
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A.(2) This event will be reviewed with (plant) personnel involved with making reportability evaluations with emphasis on the use of engineering judgment in the process.
This will be complete by September 15, 1988.
A.(3) The above corrective actions that have been taken are sufficient to prevent recurrence.
DATE HHEN FULL COMPLIANCE HILL BE ACHIEVED A.(1) Aspects addressed by the Notice of Violation will be reported in an LER by September 30, 1988.
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i A.(2) ?spects addressed by the Notice of Violation will be reported in an LER by September 30, 1988.
A.(3) Full compliance was achieved with the submittal of LER 88-008 on April 29, 1988.
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10 CFR 50, Appendix A, Criterion 2 and 44, requires, in part, that systems important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, and further requires that the design bases for these systems reflect appropriate combinations of the
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effects of accident conditions with the effects of the natural phenomena.
Contrary to the above, at the time of the inspection. (1) the design of
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the valve motor operator control circuits for the surge tank outlet valve of the Component Cooling Hater (CCH) system did not include analyses of adverse effects of earthquakes; and (2) the design bases of the CCH i
system did not reflect the combination of the effects of the surge tank outlet valves for both trains spuriously closing in conjunction with a
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safc shutdown earthquake.
This is a severity level IV violation (Supplement I).
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RESPONSE
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REASON FOR THE VIOLATION
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SCE agrees that 1) the design of the component cooling water (CCH)
system surge tanks outlet valves' motor operator control circuits did
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not include analyses of adverse effects of postulated earthquakes and 2) the design of the CCH system did not reflect the combination of
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the effects of the CCH surge tanks outlet valves for both trains spuriously closing in conjunction with a safe shutdown earthquake.
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The automatic c1csure of the CCH surge tanks motor operated valves (HOV)
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on low low surge tank level is not required to be a safety function.
1he function of the automatic closure of the HOV's is to prevent potential nitrogen ingress into the CCH system in the event of a significant water l
inventory loss. Because this function was not considered to be a safety
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function, the associated power and control circuits were_not designed to Class IE requirements. Additionally, the remote possibility of a common
mode failure (earthquake) of control relays in the MCC resulting in valve t
closure was not identified when the system was designed. Consequently, no evaluation of the occurrence was performed and the relays were not required to be seismically qualified.
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i CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED The power supplies have been disconnected from CCH surge tanks outlet valves' motor operators by removing the thermal overloads from the HOV breakers. Automatic, remote manual, and inadvertent actuation of the MOV's are thereby prevented.
The potential common mode failure of the
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circuitry causing both valves to be closed has been eliminated.
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CORRECTIVE STEPS THAT HILL BE TAKEN TO AVOID FURTHER VIOLATION
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The CCH system will be reviewed and analyzed in detail and appropriate
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action, if any, will be taken to comply with the requirements of l
10CFR50, Appendix A Criterion 2 and 44.
This evaluation will occur f
concurrently with the completion of the second phase of SCE's CCH
operability assessment discussed in our June 24, 1988 letter.
This
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i activity is expected to be completed by November 30, 1988.
In addition I
to this near term review which is focused on the CCH System SCE will i
undertake a broader review of non seismic controls for potential adverse
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impact on other safety functions.
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DATE WHEN FULL COMPLIANCE HILL BE ACHIEVED
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Full compliance was achieved for both Units 2 and 3 on June 17, 1988 when the power supplies were disconnected from the motor operators as
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i C.
San Onofre Nuclear Generating Station (SONGS) Technical Specification 4.0.5 requires in part that inservice testing of ASME Code Class 1, 2,
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i and 3 components shall be performed in accordance with Section XI of the
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ASHE Boiler and Pressure Vessel Code (Code),
The ASHE Code,Section XI requires, in part, that Category B valves be
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exercised every 3 months and that a record of test results be maintained.
Contrary to the above, as of June 10, 1988, Salt Water Cooling system
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valves HV6494 and HV6496 were not included in the Unit 2/3 inservice testing program.
This is a severity level IV violation (Supplement I).
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RESPONSE REASONS FOR THE VIOLATION SCE admits that the saltwater overboard valves (from the component cooling water heat exchangers) HV6494 and HV6496 were not included in the ASME Section XI. Inservice Testing (IST) Program.
In the development of the IST Program, valves HV6494 and HV6496 were originally considered passive, rather than active type valves, based on the fact that they have no automatic function and receive no signal from the Engineered Safety Features Actuation System. Category B passive valves are exempted from inservice testing under the provisions of ASME Boiler and Pressure Vessel Code,Section XI Table IHV-3700-1.
He have reexamined our original evaluation of HV6494 and HV6496 with regard to ASME Section XI criteria. A'more conservative interpretation
of what constitutes an active or passive valve would require that these valves be included in the IST program as Code Class 3. Category B, active valves.
This determination is based on the requirement for the valves to be manually opened following a seismic event that disables the normal salt water cooling system discharge path from the component cooling water heat exchangers.
CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED HV6494 and HV6496 have bean added to the IST program and will be manually cycled on a quarterly bases.
In addition, a Position Indication Test will be performed at each refueling outage as required by the code.
The procedure entitled Inservice Testing of Valves Program 5023-V-3.S.
i was revised to include HV6494 and HV6496 on August 18, 1988. Operations procedure, Inservice Valve Testing, Quarterly, 5023-3-3.30, was revised on August 19, 1988, to include the valves.
In3ervice Testing of the Saltwater Overboard Valves was satisfactorily
accomplished on August 20, 1988 for both Units 2 and 3.
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CORRECTIVE STEPS TAKEN TO AVOID FURTHER VIOLATIONS SCE will perform a review to determine if additional manually operated I
valves, required to bring the plant to cold shutdown, were inadvertently omitted from the IST Program.
This review will be completed by December 31, 1988.
DATE WHEN FULL COMPLIANCE WAS ACHIEVED i
J Full compliance was achieved on August 20, 1988 when HV6494 and HV6496 l
were incorporated into SCE's IST program and satisfactorily tested.
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ENCLOSURE 2 REPLY TO A NOTICE OF DEVIATION Appendix B to Mr. Martin's letter dated August 3, 1988 states in part:
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The updated San Onofre 2&3 FSAR, section 9.2.2, Component Cooling Hater System, paragraph 9.2.2.1, Design Bases, states in part:
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The component cooling water system is designed to provide a radiation monitored intermediate barrier between the reactor auxiliary systems fluid and the saltwater cooling system during nonaccident conditions.
Paragraph 9.2.2.2.1 states in part:
"The system is continuously monitored for radioactivity and all components can be isolated."
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"Radioactivity levels in the noncritical loop return header are continuously monitored in the control room to indicate any leakage of radioactive fluid into the component cooling water system.
Paragraph 9.2.2.2.3.2, Normal O?eration, states in part:
"Ouring normal system operation, one redundant loop consisting of one component cooling water pump, one component cooling water heat exchanger, and one saltwater pump is in service supplying cooling water to the various components in the nontritical loop and to critical loop A. Critical loop B is in wet standby...."
Contrary to the above, the Component Cooling Hater systems are currently and have, since the startup of Unit 2, beer, operated in accordance with 5023-2-17, Component Cooling Hater Pump and System Operation, with both loops running. The monitored nontritical loop being supplied from one loop and the letdown heat exchanger being supplied from the other. This mode of operation provides no monitoring for the loop containing the letdown heat exchanger and an improperly located sampling point for the
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loop that is monitored.
This is a deviation.
RESPONSE REASON FOR THE DEVIATION
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The deviation initially resulted from a failure to adequately evaluate FSAR requirements during procedure development and was perpetuated by the failure of existing mechanisms to identify and capture changes to the FSAR. Although CCH operating philosophy was examined on at least three separate occasions prior to the SSFI, either the deviation was not identified or was identified and not resolved.
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During startup testing in 1981, both CCH loops were operated to facilitate the scheduling of testing of equipment supported by cach loop and to avoid potential damage to equipment which could occur if it was started and the associated CCH loop was not operating. After startup, this operating philosophy was adopted from the startup procedures and continued by operating procedure S023-2-17. The deviation from the FSAR was not recognized when the startup procedure was adopted and the FSAR was not changed accordingly.
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In late 1982, an operator error rendered the train A emergency chiller inoperable due to the chiller being aligned to an inoperable CCH train.
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In response to an Onsite Review Committee (OSRC) request related to this l
incident, the Nuclear Safety Group (NSG) evaluated the capability of the emergency chillers to function during transients at one unit while
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aligned to the other unit.
This evaluation resulted in a recommendation for strict adherence to the then normal CCH System operating practice of two loop CCH operation to maintain high reliability for automatic start of the Emergency Chilled Hater System (ECHS), and a design change to provide for starting of all CCH pumps in both units upon actuation of the
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emergency chillers, which would allow a return to one train CCW operation, consistent with the FSAR.
The deviation of two loop CCH System operation from FSAR was recognized but action was not taken to
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update the FSAR.
The design change was not implemented because it was obviated by the established CCH System two loop operating practice.
Although SCE had identified the general need for FSAR changes for two loop CCH operation, the design changes proposed would have made the FSAR
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There was no mechanism to ensure that the FSAR i
changes would have been made following cancellation of the design change.
In addition, the radiation monitoring deviation was not recognized as such, at this time, since the NSG considered routine sampling of the CCH System to be consistent with the FSAR.
j In 1986, as a result of a productivity improvement program suggestion to i
operate the CCH system in accordance with the original design to avoid
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excessive equipment wear, an analysis was conducted to compare two loop
operation versus one loop operation for the CCH system.
The analysis
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identified the radiation monitoring deficiency with two loops operating,
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for the first time.
The radiation monitoring deficiency was noted but no
action was initiated to rescIve the deviation from the FSAR.
l As a result of the SSFI, NCR-G-0867 was issued which identified the
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discrepancy between the single loop operating design basis and the actual two loop normal operation as well as the deficiency in the radiation
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monitoring alignment.
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CORRECTIVE STEPS TAKEN TO AVOID FURTHER DEVIATIONS i
The operating procedure was revised on June 17, 1988 to align the letdown heat exchanger on the same CCH loop as the non-critical loop, resolving
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the radiation monitoring problem.
The FSAR will be revised to reflect
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the current operating practice.
This will be completed when the next annual FSAR update is submitted which is due by February 16, 1989.
Further corrective actions will be identified with SCE's action plan to a#ress the basic issues identified by the SSFI team which will be forwarded by October 3, 1988.
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HEGION V
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Southern California Edison Company
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33 0CT 4 A8: 52 2744 WALNUT 1ROvt AVENUC
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October 3, 1988
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U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region V 1450 Haria Lane, Suite 210 Halnut Creek, CA 94596-5368 Attention: Mr. John B. Hartin, Regional Administrator Subject: Docket Nos. 50-206, 50-361 and 50-362 Independent Assessment of Engineering and Technical Support San Onofre Nuclear Generating Station Units 1, 2 and 3 TN,;rpose of this letter is to discuss the results of SCE's assessment of technical support for San Onofre Nuclear Generating Station and ongoing and planned future actions to restore confidence in SCE's control of technical work.
Over the past year you have expressed increasing concern about the adequacy of SCE's control of technical work.
This cancern results from the lapses in the Unit 1 EQ program, incorrect diesel generator calculations, use of incorrect battery load profiles ir. surveillance tests, failure to adequately translate design requiremants into operating procedures, and inadequate command of the CCH system design bases which was evident in the recent Units 2 and 3 SSFI.
These problems are of serious concern to SCE in that they are indicative of a breakdown of SCE's control of engineering and technical work and do not reflect the high standard of engineering excellence
'which SCE must achieve and maintain to inspire continued confidence in the nuclear option. As was discussed in the June 10, 1988 SSFI exit meeting and in our responses to the Notices of Violation related to the SSFI and Unit 1 EQ, SCE committed to an exhaustive independent assessment of technical support for the operation of SONGS.
This letter provides an overview of the findings of this assessment and the actions which will be taken to achieve a high standard of engineering excellence.
SCE established a Task Force to perform an Independent Assessment of Engineering and Technical Support for San Onofre Nuclear Generating Station.
The Task Force was chartered with determining the root causes of the apparent breakdown in SCE's control of technical work, and recommending specific corrective actions.
drawn from a number of different organizations.The Task Force was composed of In excess of 10,000 man-hours 22dGldGdi7 D
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J. B. Hartin-2-October 3, 1988 were dedicated by Task Force personnel in the past three months to the assessment of engineering and technical support.
A three volume page report documents the Task Force activities including over on,e hundredtwo tho specific recomendations.
the principal conclusions and recomendations is attached.A summary of The Task Force concluded that the major contributors to the existing proble are the complexity of the current organization, heavy reliance on engin contractors combined with inadequate allocation of SCE engineering reso and the lack of readily accessible design basis documentation.
i actions to address these conclusions, the Task Force has recomended aAs correc reorganization with responsibility for design functions and the design basis focused in one department, the augmentation of in-house engineering resour and performance of all conceptual engineering in-house, and the establishment of a design basis documentation (DBD) program to recapture and maintain the design basis for all three units.
recommendations to resolve the identified problems.SCE is committed to act on th l
As an initial step in implementing the Task Force recomendations, SCE's nuclear related activities formerly distributed among three separate of the Fuel and Haterial Management Department), hav effective today, under two departments.
Essentially, the reorganization transfers the design and engineering nuclear support functions provided b Safety and Licensing (NES&L) Department. Engineering and Constr The effect of this will be to focus design basis into one organization.the responsibility for all design-relat
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utilize engineering resources and simplify the interface between the of
nuclear support functions and the Nuclear Generation Site (NGS) Department.
In addition NES&L will move from the Rosemead General Office to Irvine,
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significantly reducing the distance to San Onofre.
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site will result in improved inter-department comunications and willThe close proxim(
facilitatt increased cross training between NES&L and NGS personnel.
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' department will result in more efficient utilization of exis
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effectively increasing the resource devoted to engineering activities.
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new organization, SCE will increase the level of in-house In the l
by:
enhanced technical training necessary to perform their functions, and 3) by increasing the level of supervisory involvement in the design process.
increased level of in-house technical expertise resulting from these actions The will permit SCE to reduce reliance on engineering contractors and increase the amount of conceptual engineering performed in-house, over a period of time as
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in-house expertise 1 developed.
SCE intends, eventually, to perform all
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J. B. Martin-3'-
October 3. 1988 conceptual engineerin3 and some detal. led design in-house.
Recognizing that, I
in the near term at least, some degree of conceptual engineering and the majority of detailed design will continue to be contracted, SCE will negotiate long term engineering contracts with a limited number of contractors under which SCE could effectively exercise control over lead contractor personnel to ensure continuity of expertise.
A significant feature of the new organization is the establishment of a Design Basis Documentation (DBD) Organization.
The Design Basis Documentation Organization will be responsible for the overall development and maintenance of comprehensive design basis documentation.
SCE will primarily utilize in-house resources for DBD development.
This, in combination with extensive in-house inter-disciplinary review, will foster a high degree of famillerity with the design basis documentation by the conclusion of the DBD development e tfort.
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The SSFI noted that the SONGS 2/3 FSAR contained numerous discrepancies.
SCE will utilize the Design Basis Documentation development effort to identify existing discrepancies which will be corrected in annual FSAR updates.
SCE will improve the FSAR change process to more effectively identify the need for changes to the FSAR and will reduce the scope of the FSAR by removing repetition and information which is no longer relevant.
As noted above, an extensivt investigation, findings and recommendations is available. report documenting the detl t
long and complex and this letter only provides an overview of the mostIt is necessarily
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eignificant. conclusions and recommendations.
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with you in the near future to discuss the detailed findings and specificSCE will i
recommendations, and to answer any questions you may have regarding the
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actions we are taking.
The reorganization and relocation represent a major change which will foster the development of a high degree of in-house technical expertise and design engineering c.apability.
Because of the magnitude and significance of these
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changes, we anticipate that it will be some time before the new organization coalesces and develops its full potential.
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Now that the initial step has been taken with our new organization becoming effective today, our managers will now turn their attention toward implementing the specific corrective actions in their areas of responsibility.
SCE will strive to implement the corrective actions and achieve our goals as soon as is practical.
SCE will keep you informed of the status and schedule for implementation of corrective actions
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as they are developed.
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be provided within three months.As a minimum, a schedule for the DBD development will l
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J. B. Martin-4-October 3, 198'8
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SCE is confident that the actions we are taking to implement the Task Force i
recommendations will be effective in alleviating NRC concerns regarding SCE's f
coptrol.pf,techn{calwork._
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If you have any questions regarding the actions we are taking, please call me.
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Very truly yours, I
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Enclosure
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U. S. Nuclear Regulatory Commission Document Control Desk F. R. Huey, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3
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,,ECLOSURE
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INDEPENDENT ASSESSMENT OF ENGINEERING AND TECHrtICAL SUPPORT FOR SAN ONOFRE NUCLEAR GENERATING STATION SCE established a Task Force to perform an Independent Assessment of
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Engineering and Technical Support for the San Onofre Nuclear Generating Station as a result of the increasing emphasis on engineering excellence within the nuclear industry, and in response to NRC Safety System Functional
Inspection (SSFI) findings at San Onofre.
The purpose of the Independent
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Assessment was to:
1) identify programmatic deficiencies in engineering and technical sup deficiencies, port to San Onofre and determine the causes of those1
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to organizational structure and staffing levels, and 111) to compare the SSFI findings at San Onofre with the findings and subsequent commitments of other utilities, and recommend specific corrective actions based on those findings.
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The analysis was divided into the following areas:
l Root Cause Analysis i
o Documentation and Audits Review: Review of pertinent documentation of plant probles.t to identify pro
necessary corrective actions. grammatic deficiencies and to identify
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o Root Cause Evaluation: Detailed evaluation of seven engineering related problems to determine root causes, Comparison of Task Force Corrective Actions, SSFI findings, and o
corrective actions previously identified by SCE.
i Resource and SSFI Commitment Analysis i
o Resource Analysis: An analysis comparing SCE's organizational structure l
and the engineering and technical support staffing levels with those of
the Region V and five other selected utilities.
r o SSFI Commitments Analysis: A review of the SSFI findings and subsequent commitments of those selected utilities that have undergone an SSFI to
assess similarities between their SSFI findings and those of SCE.
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The results of the above tasks were then reviewed to identify the actions
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necessary to enhance the quality of engineering and technical support at San i
Onofre.
The principal conclusions and recommendations resulting from the
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assessment can be summarized as they relate to Organization, Resources and
i Design Basis Documentation:
l ORGANI1A110N Conclusions:
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The organization of three nuclear-related departments anj the current division of responsibilities can be improved.
While the current situation
was adequate for the time period immediately following commercial operation of Units 2 T. 3, a consolidation of nuclear functions and focusing of design responsibilities is now appropriate. A contributing cause to several of the identified problems was the complexity of the current organization, the fragmented design engineering responsibility, and resulting interorganizational communication difficulties. As a result, no single organization was charged with the responsibility and accountability for maintaining both the design and the design basis.
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Recommendations:
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Reorcanize Into Two Nuclea'r Deoartments
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The nuclear functions should be reorganized from three departments into'
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two departments and the design functions and responsibilities should be
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focused in one department.
This consolidation would improve interface
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between organizations.
In addition, consolidating design engineering
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functions within one department and allocation of resources resu(NES&L) will improve the prioritization
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engineering and technical personnel.lting in a more efficient utilization of (
consolidation of nuclear functions include:Other reasons supporting l
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The quality and efficiency of engineering activities can be increased
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by reducing duplication of effort and further clarifying-
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accountability.
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i Hanagement attention will be dedicated to and focused on nuclear
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matters without distraction by non-nuclear activities.
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Further consolidating design engineering functions within the NES&L i
o Department and systems cognizant engineering functions within the NGS Department will result in a separation of "design" and "systems"
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engineering functions as recommended by INPO.
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Virtually all the utilities surveyed have consolidated design o
engineering functions to improve communications and further clarify
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accountability.
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RESOURCES
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Conclusions:
Engineering contractors no longer maintain an adequate level of nuclear i
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plant-specific expertise and knowledge.
This is in part representative i
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accordingly, engineering contractors no longer have contractual
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incentives to maintain the staff and training programs necessary to
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ensure adequate technical resources for a specific nuclear plant.
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also does not have adequate resources to properly support all required
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I routine engineering activities and acequately supervise engineering
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contractors. This insufficient oversight of contractors by SCE has aggravated or caused difficulties in a number of the identified l
engineering deficiencies. Under the current conditions where contractor
expertise has diminished, additional SCE overview and supervision is
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required.
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I Recommendations:
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i Devote More Resources to Design-Related Enaineerino
SCE should increase the staff devoted to design-related engineering by:
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i) transferring engineers that are performing design-related engineering i
in other departments to the one group that will be responsible for
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design,11) adding staff in key technical skills areas,111) providing
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personnel with the technical and procedural training necessary to j
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adequately perform their functions and iv) by increasing the level of l
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supervisory involvement in the design process.
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Perform Concentual Enaineerina In-House
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SCE should perform the majority of conceptual engineering in-house and evaluate performing additional final design activities in-house.
Performing conceptual engineering in-house facilitates a better in-house understanding of the plant design and is responsive to the NRC's position with respect to assuming more responsibility for design work.
This permits SCE to increase the level of engineering excellence and allows for' in-house control of the key areas of design-related engineering that affect the design basis and design integrity of the plant.
This will further develop and help maintain design and design basis expertise that cannot otherwise be achieved.
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Neactiate Lono-Term Enaineerina Service Contracts
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Long term engineering service contracts should be negotiated with a limited number of contractors.
This allows the contractors to efficiently maintain technical competency and allows SCE to specify and obtain control of lead engineering contractor personnel to ensure a continuation of contractor expertise.
DESIGN BASIS DOCUMENTATION Conclusions:
A recurring contributing cause of the problems related to engineering and technical support is that the design basis of San Onofre 1, 2 and 3 is documented in a variety of sources that were not known to or not readily identifiable and/or accessible to the personnel involved in the design process.
No programmatic requirements exist requiring the update and compilation of the Unit 1 Design Criteria documents as criteria is modified or created for a particular design change except when these changes affect an ongoing retrofit project. Design criteria have not
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always accurately reflected operational and/or licensing requirements which has led to improper criteria and thus improper designs.
Therefore, it has been difficult for engineering personnel to consistently and accurately determine the design basis of a given system and convert this design basis into the implementation design criteria necessary to develop a design change.
Some design affecting documents (e.g. special studies) did not always receive a formal or completely documented interdisciplinary review.
The scope of each reviewers responsibility is not clearly defined during interdisciplinary reviews.
Further, it has been difficult to ensure that all of the design basis considerations are included in the development and maintenance of station operating procedures.
A primary source for design basis information has been the Final Safety Analysis Report (FSAR), however, this document is now considered inappropriate for use as a design basis by the NRC and INPO.
Recommendations:
Initiate a Desian Basis Documentation Program SCE should develop an integrated design basis documentation program for San Onofre Units 1, 2 & 3.
This responds to NRC concerns germane to the ability of SCE engineering and technical personnel to control the design and demonstrate understanding of the plant's design basis.
It will further improve the quality of design work through improved
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accurac basis. y, accessibility, definition and understanding of the design
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plant design basis.It will further establish utility knowledge and control of the
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the physical plant configuration and the plant design basis.And, it w It provides a tool for future design basis training for SCE engineering and technical staff, and will facilitate the response to questions regarding impact of new regulatory issues on the plant's design basis.
The program should address the following:
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The documentation should provide access to the complete design basis o
by all personnel involved in the design process, and the
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documentation should include or reference all pertinent engineering
and licensing documents.
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The documentation must be administratively controlled by procedures o
that ensure that all information that could effect the design basis (such as licensing, operations and design changes) is properly reviewed, that appropriate information is incorporated, and that the documentation is current.
The process of formulating design basis documentation will educate personnel on the design basis.
Once assembled, the documentation then becomes an invaluable tool for training on plant design and design basis.
These conclusions and recommendations reflect the changing nature of SCE's
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responsibilities in operating the San Onofre Nuclear Generating Station.
San Onofre Unitt 2 and 3 began commercial operation in August 1983 and April 1984, respectively, endin approximately ten years. g a successful period of construction that lasted completed the transition from a construction to an operating site.Over the The operating environment has different challenges from those which existed
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during the construction of Units 2 and 3.
For example, the role of engineering contractors has reduced significantly in recent years with a corresponding reduction in specific nuclear design-related expertise.
As a result, SCE must take action to ensure that adequate design-related expertise is devoted to the design process.
The recommended actions are intended to refocus and augment SCE's nuclear organization so that responsibilities are more centralized, interfaces stepitfled, and the level of design-related expertise is increased.
This, along with programmatic changes such as initiating a Design Basis Documentation program and increasing the level of in-bouse design work, will enhance the quality of engineering and technical support at San Onofre.
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