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July 29, 1970 J. P. O'Reilly, Chidf, Reactor Inspection & Enforcement Br.,
Division of Casspliance, Headquarters
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JERSEY CENTRAL POWER & LIGHT CCHPANY (0YSTER CREEK 1)
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DOCKET NO. 50-219 The attached report of inspection trips to the. subject utilities manag-ent
-offices in Parsippany, N.
J., the Gen'eral Public Utilities engineering offices in Parsippany, N. J.,
and the subject reactor site during the period May 16-22, 1970, is forwarded for action.
Six items of noncompliance were identified. An additional five items _
y considered worthy of DEL's attention are spoken to.
Several other matterc of significance are also discussed.
g.4 The items of noncompliance are as follows:
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1.
Failure to meet a limiting condition for operation pertaining to major ET D
steam break protection on January 1,1970.
2.
Failure to provide prompt notification and submit a written report on item 1..
3.
Due to the loss of automatic relief protection of two (of four) reactor g
coolant relief valves during plant operation on March 25, 1970, allimiting safety system setting was violated (in effect).
4.
Failure to include the significant aspects discussed in item 3. in reporting to the Commission on the event related thereto.
5.
Failure to exercise control rods, on April 15, 1970, in accordance with technical specification requirements.
6.
Failure to submit, in a timely mannar, the written report of the most recent (April 21, 1970) failure of the main steam isolation valves to meet specified leakage limits.
The items ud roccumnend be brought to the attention of DEL for their considera-tion are ac follows:
1.
The leak rate performance history on the main steam isolation valves is unacceptable. We strongly urge that DEL require of this licensee, and such others as is appropriate, a more frequent testing schedule, i.e.,
once each four months vs each refueling outage, until acceptable performance 9508070208 950227 C 0 M P LI A F CE omcr >
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of this important component in the primary coolant boundary had been demonstrated.
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2.
N two incidents favolving the loss of protective instrumentation high-light the potential for undetected less of important instrumentation thet could result frasa a olosure of an essess flow check valve..This in turn raises serious questions regarding the application of these valvs.s in nuclear plants. We re==aad that DRL request CE (generic l
eor:.siderations) to justify the design adequacy of these check valves.
3.
N soebers of the type identified as betag installed in 0C-1 (scintered i
metal) are considered to have a potential for increasing instrument I
response time (crud pluggage). Cdasequently, we esasider that the use of saubbers or any devienthat has.the potential for increasing protective 1
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system response times should be covered in technical specifications, i.e., requirements that adequate instrument response times.be periodically
~4 confirmed. We reconnend that DRL give consideration to this matter.
4.
In ancordance wit 4 prior telecons with DP1, we do not recomunend the approval of.the JC submittal dated June 30, 1970, concerning the release of Mr. Hatrick, Operations Supervisor, front direct operating responsi-bilities on August 1, 1970. At present there are three licensed shift' foremen at the site and therefore we consider the release of Mr. Hatrick at this time not to be in the best interests of safety.
i 5.
h JC review of the March 25, 1970, incident involving the disconnecting of a primary system instnanent sensing line, disclosed that a design error was responsible for the loss of automatic Isolation Condenser actuation from a high reactor pressure. We consider that this design error may be i
generic in nature and should be reviewed for other boiling water reactors.
other matters of significance include the following:
1.
W assigned inspector h.s advised JC that more prudent pipe wall thickness measurements on the north core spray nossle safe end are in order. We will pursue this matter during the next inspection and if JC does not subscribe to this position, we will forward our raea-a-dations to Headquarters.
i 2.
W issue of the Isolation Condenser instnamentation trip settings being i
set in noncompliance which was discussed in Inquiry Memoranden 219/70-H, will be reviewed during the next inspection and appropriate enforcement action taken at that time. This matter was identified during a telecon.
with the site to discuss the applicability to OC-1 of the Dresden 2
.l incident that securred on June 5, 1970.
3.
h subject of transporting concrete through al-in pipe was previously.
discussed with JC. They have informed us that this nothod was used to transport concrete to an area under the drywell inside the drywell support 7
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-3' skirt.. JC and Burns and Roe have evaluated this matter and concluded that there does not seem to be any failure mode possible.
Sampling-of I
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the concrete in question is considered impractical by the licensee as it is surrounded by e 5 feet of other high strength concrete. We have informally. reviewed this item within the region and have not identified any major eencarns.
l 4.
Several itens were identified for further review and continued special attention during Mr.. Pomeroy's assist inspection of the major overhaul of the rod drive assemblies. We plan to followup on these items which includes (a) reviewing the reassembly reports that were cont =n N ted and unavailable for review during the inspostion, (b) reviewing the results of the repeat friction tests for.four drives whose original-tests indicated marginal conformance, (6) the res61ts of pressurised scram and stall flow
.I test data, and (d) the results of the continuing surveillance by JC of-the scram times, buffer action, and southly stall flow tests.
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5.
The 242 foot elevation stack sampling ec,uipment was reported by-Mr. McCluskey to have been installed in early July, 1970 Good agree-ment was noted between' data from this source and the permanent stack 4
sampler.
6.
Per your May 7,1970 memoranden concerning the first 00-1 semi-annual report, Mr. McCluskey has been advised to submit an addendum to fully document the problems experienced and the repairs made to the control rod
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drive assemblies in October,1969.
Mr. McCluskey.has stated an addendum will be submitted, 7.
PORC's past perfomance was discussed during the exit interview with Mr. McCluskey. We considered him to be receptive and expect improved
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performance to result from these discussions.
8.
GORB's performance was reviewed with Mr. Hirst, GORB Chairman.' Observed deficiencies were discussed as were C0tI expectations of future performances.
We will reconsider the need to revisit the corporate offices to discuss this matter if improvement is not evident enring the ment inspection.
9.
Scran Hos. 42 and 43 resulted from problems with the turbine stoma pressure controller. JC has requested CE to review the design of this central system to determine if any modifications are warranted.
- 10. The information provided in the June, 1970, monthly report concernin8 Mr. Ritter's retirement was incorrect. It was Mr. Logan who elected the r
option of early retirement and who was repleted by Mr. Sims. Mr. Ritter, who was en an estended leave of abeense, has returned and is currently assigned to "special projects" reporting to Mr. Sims.
- 11. - Mr. McCluskey informed us during a telecon on July 15, 1970, that in addi-tion to the single rod drive previously identified to CO, two additional rod drives h d measured stall flows in excess of 5 spa on June 25, 1970.
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-4 The timeliness of the licensee's somewhat tardy verbal report was discussed with Mr. McCluskey and he stated that C0:1 will be prom $tly informed of all such future findings. Accordingly, we consider this matter to have been satisfactority resolved at this time end do not reemamend that this issue be included in the pending enforcasesnt letter.
Mr. McCluskey informed us during a telecon on July 20,'1970, that all rod drive stall flows were presently < 5 spa. These were reported to have been corrected by maintanar:e on the " directional control valves" for the rods in question.
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Total stall flows were reported to have increased from 167 spa on May 23, 1970 to 199 spa on July 20, 1970. We are closely following rod drive performance and will keep you inforised of significant developmente.
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Ue recotraend that the pending enforcement letter be addressed to Mr. R. F..
1 Bovier, JC President (cc to W. G. Ruhns) and updated to include the previous j
two inspections on March 10-20 ar.d April 21-22, 1970, and include the sin items of noncompliance that were identified during the most recent inspection on j
May 18-22, 1970.
In addition to the items outlined in the draf t fom of the i
letter, we consider that JC should address themselves to the adequacy of the i
presently existing corporate management audit system and personnel staffing of I
the OC-1 facility for assuring the safety of operations.
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=1 R. T. Carlson Senior Reactor inspector f
i Attacleient:
CO Report lio. 219/70-5 by R. J. licDemott, dated 7/29/70 (Mato) i I
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