ML20214Q875
| ML20214Q875 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/19/1986 |
| From: | Frisch R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20214Q878 | List: |
| References | |
| NUDOCS 8609290001 | |
| Download: ML20214Q875 (5) | |
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j/ Company Big Rock Point Nuclear Plant, Box 591, Route 3, Charlevoix, MI 49720 September 19, 1986
- Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - JUSTIFICATION FOR INTERIM OPERATION FOR BACKUP CONTAINMENT SPRAY VALVE MO-7068 During the NRC audit conducted the week of September 15, 1986 of Big Rock Point Plant's compliance to 13 CFR 50.49 Environmental Qualification (EQ) it was discovered that uncertainties exist in the qualification status of the plant's backup containment spray valve actuator.
The purpose of this letter is threefold:
- 1) to describe the events that culminated in our conclusion that the qualification is in question, 2) to identify our immediate actions taken to justify actuator operability for the short term, and 3) to describe our planned longer term actions to effect full qualification.
Following the April, 1975 simulated LOCA testing of M0-7068 at Franklin Research Laboratories, the actuator was shipped to Limitorque for post-test inspection and refurbishment. At Limitorque the actuator was disassembled.
No mechanical failures of any kind were observed but the following irregular-ities were noted:
1.
The grease in the actuator had lost some of its lubricity due to age and separation of the oil from the grease.
2.
Some water (approximately three to four milliliters) dripped out of the worm gear assembly area when it was removed.
3.
The zinc die cast parts of the limit switch and torque switch assemblies had a white powder (some type of corrosion) on the surface.
The actuator was reassembled and greased then returned to the plant.
At the plant, new seals were placed in the actuator and it was reinstalled in the plant because of its demonstrated capabilities.
On Janua ry 20, 1986 Engineering Planning and Management, Incorporated (EPM) requested permission from Consumers Power Company to use the Franklin Research Laboratory test report for a data package they were producing to assist c{
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t Director, Nuclear Reactor Regulation 2
Dig Rock Point Plant Justification for Interim Operation September 19, 1986 utilities in justifying the use of PVC insulated control wire.
On January 30, 1986 EPM further requested that we verify that no rewiring of the actuator was i
performed after the simulated LOCA test.
On January 31, 1986 permission was l
given to utilize the test report. We also, based on records of 1980 equipment walkdowns, verified that the MO-7068 actuator contained the PVC control wire.
Based on no documentation to support that the wire had been changed, we I
concluded that the same wire that had been tested was still installed in the actuator.
On September 15, 1986 the NRC audit team noted that the EQ file for the backup containment spray valve actuator indicated that after 13 years of installed life the actuator was subjected to LOCA testing, refurbished and shipped back to the plant where it was reinstalled and returned to service. The NRC audit team requested that Consumers Power Company provide documented evidence of the identified refurbishment.
In response, Consumers Power Company reviewed documentation associated with the 1975 qualification testing and refurbishment efforts as well as field walkdown records developed for the Big Rock Point EQ program.
On September 16, 1986 Consumers Power Company concluded that based on the documentation reviewed, the refurbishment conducted at the Limitorque facility immediately after the LOCA testing was not adequate to fully restore the actuator to an environmentally status. Replacement of all components featuring organic materials (eg, motor, control wiring) while the actuator was disassembled was not performed. After reaching the above conclusion, the NRC audit team informed and with their agreement Consumers Power Company undertook was evaluations to substantiate that the backup containment spray valve could be considered operable and be relied upon for both normal and post-accident service for a short period of time. provides a description of containment spray operation and the transients in which spray is required. Also provided is an evaluation of maximum temperature and radiation capabilities the actuator subcomponents containing organic materials can withstand. This evaluation supports our expectation that the actuator is capable of surviving another LOCA or MSLB transient and a 30 day post-accident recovery period.
Attachment I also contains a failure modes and effects analysis which shows that in the event an unexpected failure develops after some appreciable time into the accident, a failure to reclose the valve to secure post-accident spray could conceivably result af ter the valve was opened for purposes of aiding in iodine washdown. Since washdown is not a required function for the backup core spray line, the requirement to open the valve for this purpose has subsequently been deleted from existing emergency and operating procedures. provides a thermal degradation equivalency evaluation which was performed to assess the overall amount of degradation suffered by the actuator as an accumulated result of 24 years of installed life at the plant, the simulated LOCA test, and a postulated second LOCA event followed by a 30-day post-accident recovery period. The purpose of the evaluation was to compare this overall degradation and resulting loss in life to the minimum overall life expected for this actuator. Results of this weak-link, Arrhenius-based OC0986-3034A-BP01
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Director, Nuclear Reactor Regulation 3
Big Rock Point Plant Justification for Interim Operation September 19, 1986 I
evaluation show that a life of 11 years remains; a duration significantly greater than the duration to our next refueling shutdown when replacement will be effected.
In performing the attached failure modes and effects analysis and aging evaluation Consumers Power Company is assuming that the containment spray I
system is subject to the requirements of 10 CFR 50.46(d); in particular the
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single failure requirements of 10 CFR 50 Appendix A, GDC 35 are assumed to apply. This assumption is based on the following assessment of regulatory requirements:
1.
10 CFR 50.49 requires environmental qualification of electric equipment important to safety.
This equipment is that relied upon to remain functional during design basis events, which among others includes design basis accidents such as LOCAs and steam line breaks.
2.
LOCAs and steam line breaks are those design basis accidents governed by the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K.
One of the requirements endorsed by 10 CFR 50.46 is that the ECCS system function be accomplished assuming a single failure.
3.
It is assumed tuat the equipment necessary to meet the single failure requirement of 10 CFR 50.46 must also comply with the environmental qualification requirements of 10 CFR 50.49.
This equipment is a subset of that appearing in the Big Rock Point emerp cy procedures for loss of coolant accidents.
4.
In demonstrating compliance of the ECCS with the single failure criterion, two simplifying assumptions are made; first that no credit is taken for the operation of equipment that is not qualified for the environment in which it must operate, and second that if the environment should exceed the profile to which a piece of equipment is qualified then that equipment is not capable of performing its intended function.
With these assumptions as background, Consumers Power Company performed failure modes and effects analyses on the core spray system assuming that MO-7068 failure can occur in one of two manners; in the fully closed position given that it has not yet received a signal to open at the time the environment degrades, and in the fully open position assuming that it successfully opens but fails subsequent to its initial operation.
Evaluation of single failure adequacy during both the initial injection mode of core spray cooling and the long term recirculation mode of operation was performed.
The failure modes and effects analysis performed assuming the valve fails in the closed position quickly revealed a potential deficiency with respect to compliance with the single failure criteria of 10 CFR 50.46 during the initial injection mode.
That was, with MO-7068 closed prohibiting flow through the backup containment spray header, a single failure of the primary spray valve (MO-7064) following a steam line break would result in a containment environment that exceed the qualification of a large block of safety related equipment located inside the containment (all four core spray valves, fo r example).
Under assumptions made in item 4 above, adequate core cooling cannot OC0986-3034A-BP01
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Director, Nuclear Reactor Regulation 4
Big Rock Point Plant Justification for Interim Operation September 19, 1986 be assured in this ECCS configuration. The evaluation performed assuming that 1
the valve fails in the open position revealed only one minor deficiency. That had to do with the use of MO-7068 for iodine washdown during the long term recirculation phase of ECCS operation.
Given these results, it was recognized that to provide assurance that the Big Rock ECCS remained in conformance with the single failure requirements of 10 CFR 50.46 several actions with respect to the qualification and operation of MO-7068 were necessary. First because of the single combination of events associated with MO-7068 being failed closed (a steam line break with coincident failure of MO-7064 to open), it was necessary to show that operability of the valve can be assured during the first few minutes of the event permitting it to be placed in the open position prior to its failure.
Second, a deletion of the requirement to use the valve to provide iodine washdown was desirable.
The attachments contain the necessary documentation to provide the above assurance. Thermal aging calculations and a comparisen of typical motor and cable capability to withstand the accident environment are provided. These demonstrate the ability cf the actuator to survive the initial stages of the accident for a sufficient duration to place the valve in the open position should it be necessary (required only if random failure of MO-7064 occurs).
Satisfactory, long term operation of the ECCS is demonstrated by the failure modes and effects analysis performed assuming MO-7068 is failed open.
As stated in the evaluation, adequate core cooling is assured provided the need for MO-7068 for long term iodine washdown is removed from current procedures.
In summary, Consumers Power Company acknowledges that the limited refurbishment performed af ter the 1975 LOCA simulation test leaves uncertainty as to the actuator's capability to provide service for the life of the plant. Nevertheless, this evaluation provides reasonable assurance that the valve can be depended on for a shorter period of time until replacement or refurbishment of the actuator can be made at the earliest opportunity but no later than the 1987 refueling outage.
Additional testing or inspection during this time period is not considered necessary or desirable.
Plant historical records show that the valve has not experienced any failures to operate throughout its inservice life since it was LOCA tested.
Additional testing or inspection would require entry into limiting conditions of operation.
In addition, since the spray lines are normally filled, stroking the valve would require upstream manual isolation which would necessitate a concurrent removal from service of a redundant core spray line as well.
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Big Rock Point Plant Justification for Interim Operation September 19, 1986 The Big Rock Point Plant Review Committee has reviewed the evaluations and concurred that continued plant operation will not present a significant hazard to public health and safety untii MO-7068 valve actuator is replaced or refurbished.
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Ral h R Fr1 Senior Licensing Analyst CC Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point Attachments l
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