ML19336A617
| ML19336A617 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/21/1980 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19336A605 | List: |
| References | |
| 50-320-80-09, 50-320-80-9, NUDOCS 8010300323 | |
| Download: ML19336A617 (3) | |
Text
_ _ _ _ _ _ _ _ _
APPENDIX A O"
NOTICE OF VIOLATION Metropolitan Edison Company Docket No. 50-320 Based on the results of an NRC inspection conducted on April 27 -
May 31, 1980, it appears that certain of your activities were not conducted in full compliance with the conditions of your NRC Facility License No. DPR-73 as indicated beltw.
Items A, B'and C are categorized as infractions, and item D is categ;rized as a deficiency.
A.
The Order for Modification of License, dated, July 20, 1979, as amended by the Order dated February 11, 1980, states in part:
"...Pending further amendment of the Fac'ility Operating License, the licensee shall maintain the facility in accordance with the requirements set forth in Attachment 1..."
(propoa d Technical Specifications, Appendix A to License No. DPR-73).
The proposed Technical Specification 6.8.1 states in part:
" Written procedures shall be... implemented covering...the applicable procedures recommended in Appendix ' A' of Regulatory Guide 1.33, Revision 2, February 1978...."
Appendix ' A' of Regulatory Guide 1.33 lists procedures for the Plant Fire Protection Program.
Fire Protection Procedure 1410-Y-26, Revision 4, April 23, 1980, Welding, Cutting, Grinding and Open Flamework Procedure for Fire Safety, paragraph 6.1.1 states:
"A permit system shall be provided to positively control all welding and cutting in plant and buildings."
Further, paragraph 6.1.4 establishes a " permit form" (enclosure to the procedure) and this fonn requires in part:. ample portable extinguishing equipment; all combustibles have been located 30 feet to 40 feet from the operation and the remainder protected with fire retardant material; control room be informed of the operation; and, documentation of followup inspection by the designated firewatch after the operation.
Contrary to the above, on March 29, 1980, the required fire protection procedures were not implemented during open flame cutting operating in the fuel handling building, in that:
A fire protection permit system was not provided to positively control the ' cutting operation.
Ample portable extinguishing equipment was not, provided as required.
Combustible material was located five feet from the open flame cutting area.
The control room was not infonned of the operation.
No documentation of followup inspection by the designated firewatch was made after the operation.
8 010300 Kg
2 B.
The Order for Modification of License, dated July 20,1979, as amended by the Order dated February 11, 1980, states in part:
...Pending further amendme, of the Facility Operating License, the licensee shall mainta' the facility in accordance with the requirements set forth in "tachment 1...."
(proposed Technical
- Specifications, Appendix A to License DPR-73).
The proposed Technical Specification (TS) 3.6.1.1 states in part:
" Primary CONTAINMENT INTEGRITY shall be maintained and all containment penetrations, including at least two OPERABLE containment isolation valves or a double barrier in each penetration, shall be closed when not required open per procedures approved pursuant to Speci-fication 6.8.2...With one containment isolation valve per containment penetration open... maintain the affected penetration... closed with...at least one closed manual valve...."
Contrary to the above, on May 20, 1980, between 9:00 p.m. and 9:30 p.m., for apprcximately 6-10 minutes containreent integrity was not maintained at reactor building personnel airlock No. 2 with the outer door manual purge valve (designated *E" valve) and the inner door manual equalization valve open sir ultaneously.
This resulted during the implementation of Operating Procedure 2104-4.55, Revision 1, May 16, 1980, Reactor Building Entry and Pre-Decon, approved pursuant to Specification 6.8.2, which required the opening of the outer airlock door in accordance with another (referenced) procedure but neither procedure addressed the shutt!rg of the "E" valve on the outer door in the proper sequence.
s C.
The Order for Modification of License, dated, July 20, 1979, as amended by the Order dated February 11,1C03, states in part:
"...Pending further amendment of the Facility Operating License, the licensee shall maintain the facility in accordance with the requirements set forth in Attachment _1..."
(proposed Technical Specifications, Appendix A to License No. DpR-73).
The proposed Technical Specification (TS) 3.3.3.6 and TS Table 3.3-10 require in part that reactor coolant system (RCS) pressure (1 minimum operable channel) be operable.
Contrary to the above, on May 29, 1980, the only RCS pressure indications were not operable because of the inadvertent closure of a recovery system interface valve (SNS-V26) which isolated two pressure reading instruments from the RCS.
This resulted during the implementation of Special Operating Procedure R-2-80-18, dated April 3,1980, Temporary Nuclear Sample System Flowpath Verifi-cation, which utilized a valve lineup requiring SNS-V26 to be shut.
l l
i L
b
3 D.
10 CFR 50.72 states in part:
"Each licensee of a nuclear power i
. reactor licensed under...tSO.22 shall notify the NRC Operations Center es soon as possible and in all cases within one hour by telephone of the occurrence of...(3) Any event that results in the nuclear power plant not being in a controlled or expected. condition while operating or shutdown... (6) personnel error or procedural inadequacy which, during nonnal operations, anticipated operational occurrences or accident conditions, prevent or could prevent, by itself, the fulfillment of the safety function of those structures, systems and components important to safety that are needed to
... limit the release of radioact.ive material to acceptable levels or reduce the potential for such release..."
Contrary to the above, between May 20 and May 29, 1980, required reports to the NRC Operations Center were not properly made as noted below.
On May 20,1980, between 9:00 p.m. and 9:30 p.m., during the reactor building entry procedural implementation, containment integrity was not maintained in accordance with Technical Specification 3.6.1.1 for a period of approximately 5-10 minutes.
The NRC Operation Center was not notified until approximately 11:45 a.m. on May 23, 1980, exceeding the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit for notification (criterion (6) above).
On May 29,1980, a procedure was implemented and caused the isolation of the only operable source of pressure indication for the reactor coolant system (RC5).
The control room operators were unaware of this for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> during which pressure reduction evolutions occurred with the standby pressure control system isolated in one instance.
The NRC Operations Center was not notified (criterion (3) above).