ML20043B298
| ML20043B298 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/21/1990 |
| From: | Fitzpatrick E GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9005290009 | |
| Download: ML20043B298 (5) | |
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M Forked River, New Jersey 087310388 L 609 971-4000 Writer's Direct Dial Number:
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May 21,1990'
_U.S.~Huclear Regulatory Commission g
Attna Document Control Desk' Washington,'D.C.
20555 Dear Sir '
Subjects-
. Oyster Creek Nuclear Generating Station t'
Docket No. 50-219' Inspection Report 90-06 Response to Notices of Violation By latitor dated-April 20, 1990, the USNRC docketed the results of Routine
- Safety Inspection 50-219/90-06. Appendix A to that letter contained four Notices of Violation. Attachment I to this letter-provides the requisite CPU Nuclear response to each violation.
If_ any questions or comments should arise,=please contact Mr. John Rogers, oyster-Creek Licensing Engineer at 609-971-4893.
Very truly yours, E.E. Fitzp ick Vice President and Director Oyster Creek EEF/JJR jc r(JR-LTRS2)
Attachment'
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-Mr. Thomas Martin, Administrator Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 NRC Resident Inspector a s
Oyster Creek Nuclear Generating-Station i
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Mr. Alexander Dromerick U.S.-Nuclear Regulatory Commission Mail Station Pl-137 Washington, DC 20555 9005290009 900521 PDR ADOCK 05000219 O
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'!! s GPU Nuclear Corporation is a subsidiary of General Public Utikties Corporation jj;!
4.
Attachment I Violation 3 Technical Specification 3.2.B.2 requires "The Rod Worth Minimizer (RWM) shall be operable during each reactor startup until reactor power reaches 10% of rated power."
Should the RWM be inoperable before a startup is commenced or before the first. twelve rods are withdrawn, one startup during each calendar year may.be performed without the RWM provided that the second licensed operator verifies that the licensed operator at the reactor console is following the rod program and provided that a reactor engineer from the Core Engineering Group also verifies that the rod program is being followed."
Contrary to the above, the rod worth minimiser was not in service on February 15, 1990, when a reactor startup was initiated and seven control rods were withdrawn.
Additionally, neither a second licensed operator nor a reactor engineer were stationed with the purpose to verify that the rod program was being followed.
Response to Violation A Immediate corrective action was taken to halt the reactor startup, place the RWM in service by turning the RWM keylock switch to the " normal" position, and verify that the. correct sequence had been followed for the seven withdrawn control rods.-
To preclude this type of event from recurring, the incident critique report was incorporated as Required Reading for appropriate Operations personnel.
Additionally, procedural changes were made to more clearly define operability of the RWM system, and require verification of RWM keylock switch position prior to each reactor startup. Finally, a change was made to Procedure 201.1 " Approach to Critical" to require a functional test of the RWM during each reactor startup.
Full compliance was achieved on February 15, 1990 when the RWM was placed in service and the seven withdrawn control rods were verified to have been withdrawn in proper sequence.
Violation B Technical Specification 6.11 requires that procedures for personnel radiation protection be prepared consistent with the requirements of 10CFR20, and be approved, maintained and adhered to for all operations involving personnel radiation exposure.
Plcnt Procedure 9300-ADM-4000.11, " Rules for Conduct of Radiological Work," and Procedure 9300-ADM-4010.2, "ALARA Review Procedure," require the department initiating work to coordinate an ALWAA review. The procedures further require that radiolcgical engineering will perform an ALARA review for any task anticipated to accumulate 5 person-rem or more of Otal exposure.
Contrary'to the above, from about February 6, 1990, through February 15, 1990, while performing maintenance on the "A" reactor recirculation pump under Job Order No. 21359, the pump bearing was removed, measurements taken, and the bearing was replaced without this scope of the work receiving the required ALARA review. This scope of work resulted in approximately 11 person-rem of exposure.
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Resoonse to Violation B-GPUN concurs.with the violation, as clarified belcw.
Section 2.1 of Inspection Report 50-219/90-06 accurately. describes the circumstances of the job. When the previously reviewed scope of work was
' determined to need expansion, the Group Radiological Controls Supervisor (GRCS) reviewed the nature of added workscope with the cognizant Radiological Engineer
.(hE).
That review resulted in a determination that additional dose in excess of 5 rem would likely ensue. That review also determined that the protective measures evaluated and prescribed for the original work scope were fully applicable to the added work scope and that no additional protective measures were practical. This decision process was not documented using the supplemental form contained in a
procedure 9300-ADM-4010.02, "ALARA Review Procedure". Therefore, GPUN concurs that the ALARA review for the increase in work scope was not properly documented.
The radiological significance of this event is minimal.
Had the proper documentation been completed, no additional protective measures would have been prescribed or taken, and no decrease in radiation received would have occurred.
The Region I inspector concurred in this determination during the exit meeting.
To preclude this type of event from recurring ALARA Reviewers were instructed to document evaluations made to changes in existing ALARA Review workscopes regardless of whether changes in radiological controls are appropriate in order to document proof of-a further review.
Inspection Report 50-219/90-06 was issued to Radiological Controle personnel as Required Reading.
Amplifying instructions were also-added to the ALARA Review standing order which describes workscope change and the need to document subsequent ALARA Reviews performed to already existing ALARA Review documents.
Full compliance will be achieved with the completion of these actions, presently
. projected for May 31, 1990.
Violation C Tschnical Specification 3.12.C.1 requires fire protection spray and/or sprinkler systems listed in Table 3.12.2 be operable. Table 3.12.2 specifies operability of deluge systems 4A and 4B for the Cable Spreading Room (fire area op-FC-4).
Technical Specification 3.12.C:1 requires the,t when one or more of the above required spray and/or sprinkler systems are inoperable, the licensee establish a continuous fire watch, with backup fire suppression equipment, for those areas in which redundant systems or componento could be damaged.
Contrary to the above, from January 23, 1990 until February 8, 1990, fire suppression deluge system 4A was inoperable in the Cable Spreading Room which is an area containing redundant components, and a continuous fire watch was not established.
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-Lo Response to Violegion.C GPUN concurs in the violation..
A critique of this event was conducted, and it was determined that the root cause of t'nis event was a lack of understanding on the part of Operations Departnent supervisory personnel with regard to the potential implications of a trouble alarm on the control Room Fire Panel.
Senior Operations Department Management has issued guidance to the supervisory personnel directing the supervisors to declare
.I instrumentation and suppression systems inoperable when a trouble alarm is k
-received on a fire panel. Additionally, guidance was provided to ensure that maintenance requests resulting from instrumentation and suppression system trouble I
alarms be given higher priority.
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Full compliance was achieved with the issuance of the described guidance.
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violation D l
Technical Specification 6.8.1 requires that written procedures be established, I
. implemented and maintained that meet or exceed Regulatory Guide 1.33, Rev. 2.
Regulatory Guide 1.33, Rev. 2, endorses ANSI 18.7-1976, Section 5.2.2, which requires that procedures-be foll6wed.
Station Procedure 106.2.1, " Spill Procedure," requires that, if the event has contaminated a nonradioactive system, further use of the system shall be restricted until.the cause of-the contamination is identified and corrected, and the system has been decontaminated.
If it is necessary to continue operetion'of the system as contaminated, the procedure requires an immediate safety evaluation i
I to be performed in accordance with the requirements of 10CFR50.59.
Contrary to the above, on February 17, 1990, a spill from the No. 2 auxiliary boiler occurred which was identified to be radioactive. Further use of the i
normally nonradioactive Auxiliary Boiler System was not restricted, the cause of the contamination was not corrected, and the system was not decontaminated.
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safety evaluation in accordance with the requirements of 10CFR50.59 was not performed until March 22, 1990 to evaluate continued operation. As a result of this continued operation, another radioactive spill occurred on March 10 which i
resulted in a necond unplanned release of radioactive material to the environment.
Mappog3_ts Vtolation_y CPUN concurs with tha violation.
Initially, further use of No. 2 Auxiliary Boller was not restricted because the D011er was previously considered contaminated. Varying low levels of activity tspprouAmotely 1E-6 uc/ml) had been previously detected since 1986.
In January 1990, an increasing trend in activity was noted.
The source of contamination was identified to be a small leak in the Radwaste Evaporators.
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. Ac'tivity was monitored'and allowed to reach approximately 1E-3 ue/ml prior to commencing corrective actions. The limit of IE-3 ue/ml was the result of an incorrect interpretation of a Technical Specification limit for radioactive batch releases (section 3.6.B.1).
When activ!"v reached 1E-3 uc/ml, an increased frequency of boiler blowdowns was instits1.ed to lower the activity level.. Shortly
't afterward, the boiler and evaporators were taken out of service for maintenance.
.When_these systems were restarted on March 9, 1990, a step increase in activity.
occurred. As the next routine monitoring sample was-scheduled for March 14, 1990, this increase was not detected until the following day during recovery from a subsequent spill. On March 10,'1990, it was determined that the "A" Evaporator was the. source of the increase in activity.
"A" Evaporator was secured and "B"
g Evaporator was placed in service. Further corrective actions, including securing No. 2 boiler on March 14, 1990 to drain and refill were unsuccessful at lowering activity. On March 15, 1990, No. 1 boiler was placed in service and corrective actions were successful in lowering its activity level.
An effort to locate previously calculated acceptable activity levels for the auxiliary boiler was performed. On March 15, 1990, a 10CFR50.59 Determination and Safety Evaluation was initiated as the level of activity existing in the boiler l
was determined to exceed previously analyzed values as stated in memo dated June J
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17,.1980 from Stone & Webster to Jersey Central Power & Light Company (JCP&L Task l
No.80-204).
On March 22, 1990, the Safety Evaluation concluded that the activity levels in.the boiler, even when analyzed under worse case scenarios, did not exceed regulatory or design basis limits.
Long term corrective actions were initiated which evaluated the adequacy of the actions taken to address NRC Bulletin 80-10, " Contamination of Non-Radioactive System-in Result!ng Potential for Unmonitored, Uncontrolled Release of p
Radio;Ttivity to Environment". Engineering has been assigned a task to review-systems which could be potentially contaminated and the procedures controlling their operation. Appropriate action levels for activity will be calculated with requisite corrective actions proceduralized. This engineering task is presently projected to complete by October 31, 1990.
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