IR 05000219/1979008
| ML19248D181 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/19/1979 |
| From: | Caphton D, Rekito W, Tanya Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19248D177 | List: |
| References | |
| 50-219-79-08, 50-219-79-8, NUDOCS 7908130390 | |
| Download: ML19248D181 (9) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-219/79-08 Docket No.
50-219 License No. DPR-16 Priority
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Category c
Licensee:
Jersey Central Power and Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 Facility Name:
Oyster Creek Nuclear Generating Station Inspection at:
Forked River, New Jersey Inspection conduc#e j7 April 19-20,1979 Inspectors:
fhNf[D 6// 2r[77 T. H. $tsi t,11, Reactor Inspector date signed j
% 'ush k, Gla/79 W. A. Rekito, Reactor Inspector date signed a
date signed 6y fM
f Approved by:
4,i D. L. CapiltoVChief, Nuclear Support ta te' signed Section No.1, RO&NS Branch Inspedtfon Summary:
Inspection on April 19-20, 1979 (Report No. 50-219/79-08)
Areas Inspected:
Routine, unannounced inspection by regional ba*. inspectors if the containment integrated leak rate test (CILRT) report and licensee action on pretious inspecti;n findings.
The inspection involved 26 inspector-hours on site by two NRC regional based inspectors.
Results_1 No items of noncompliance or deviations were identified.
560
%0 Region I Form 12 (Rev. April 77)
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7008130 2 O
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DETAILS 1.
Persons Contacted
- J. Carroll, Station Superintendent
- K. Fickeissen, Technical Engineer
- A. Rone, Technical Supervisor
- D. Ross, Manager, Nuclear Generation The inspector also talked with and interviewed other members of the engi-neering and operating staffs.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findinas (Closed) Unresolved Item (219/77-05-01): Acceotance criteria for Integrated and Local Leak Rate Testing, as contained in the Technical Specifications and licensee procedures, are stated correctly.
This item is considered resolved.
(Closed) Unresolved Item (219/77-05-02): Valves V-16-30, 76 and 84 have been added to 1he licensee's request for exemption from 10 CFR 50 Appendix J testing requirements.
This item is considered resolved.
(Closed) Unresolved Item (219/77-05-03): The licensee's procedure for per-forming the Integrated Leak Rate Test has been revised.
Drywell pressure switches are not isolated from the test volume and are therefore subiected to test pressure.
This item is resolved.
(Closed) Unresolved Item (219/77-05-04):
A licensee's representative stated that Containment Isolation Valves V-23-14 and 16 are globe valves and that reverse direction local test pressure is under the stem.
This method of reverse direction testing is considered conservative.
This item is resolved.
(0 pen) Unresolved Item (219/78-34-01):
A licensee's representative stated that instrument weighting f actors would be incorporated into the CILRT pro-cedure, and administrative controls would be developed to ensure computer program changes are properly reviewed prior to the next CILRT.
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3.
Containment Integrated Leak Pata Test _
a.
General The following documents were reviewed by the inspector.
" Primary Containment Leak Rate Test," Procedure No. 666.5.007,
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Revision 0, dated November 16, 1978.
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Summary Technical Report " Reactor Containment Building Integrated Leak Rate Test," dated November 1978.
The aoove references were reviewed for accurate data transcription, proper implementation of procedural changes and correctness of the analytical methc'ls.
No items of noncompliance were identified.
b.
Satisfactory CI!21 A satisfactory CILRT was completed November 25, 1978, after an initial test failure which occurred during the period November 20-23, 1978.
The test failure is documented in IE Inspection Report 50-219/
78-34.
Due to the failure of the first test the licensee plans to conduct a CILRT in twelve plus or minus eight months from the date of completion of the successful test, as required by present technical specifications.
For the successful test, the inspector performed calculations using raw data and independently verified:
the proper inclusion of instru-ment calibration corrections, the computed air mass values, the con-tainment leakage rate and the 95~ confidence intervals.
The test L
results are presented in Table 1.
The inspector had no further questions concerning the CILRT at this time.
4.
Local Leak Rate Testing (LLRT)
a.
General The following documents were reviewed by the inspectors.
" Local Leak Rate Tests," Procedure No. 665.5.C>06, Revision 0,
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dated April 22, 1977.
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"Drywell Airlock Leak Rate Test," Procedure No. 665.5.005, Revision 1, dated September 9, 1977.
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'" Main Steam Isolation Valve Leak Rate Test," Procedure No.
665.5.003, Revision 0, dated March 29, 1977.
The inspector reviewed completed copies of these procedures from 1976 to the present and with the exception of the below items had no further questions at this time.
b.
Type B and C Testing The following items refer to the " Local Leak Rate Tests" procedure and are collectively designated as unresolved item (219/79-08-01).
(1)
Individual valve or containment penetration tests within the procedure do not include valve lineups.
Without specified valve lineups there is no assurance that downstream piping is vented, that liquid filled systems are drained or that test pressure is being applied in the same direction as accident pressure.
(?) Several data sheets and other pages containing procecural guidance, written on assorted types of paper not part of the approved procedure,have been added to the procedure.
These additional sheets constitute a change to the procedure but. do not appear to conform to procedure change requirements (e.g.
Data sheets 2, 21, 24-27 and 28).
(3)
The procedure requires that a correction factor 0.01 SLM (standard liters per minute) be added to the leak rate value read on the leak rate monitor.
It is unclear on all data sheets whether this value has been added or not.
(4) A data sheet for testing the Reactor Water Sample System (V-24-29 and 30) is part of the procedure, although a licensee's representative has stated that the system is not tes ta ble.
For the last two years of testing the data sheet has been crossed out.
This method of changing the approved procedure does not appear to confona to procedure change requirements.
(5)
For the tests conducted in 1978, Data Sheet 28 has not been signed by the Technical Supervisor as required by the procedure.
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c.
Drywell Airlock Testing The following items refer to the "Drywell Airlock Leak Rate Test" procedure and are collectively designated as unresolved item (219/79-08-02).
.(1 )
The procedure calls for allowing temperature to stabilize after pressurizing the airlock but the temperature is never measured.
(2) While taking pressure decay data during the test, the tempera-ture of the airlock is not measured.
The effect of temperature changes on the pressure decay can therefore not be determined.
A temperature rise would mask leakage and result in a noncon-servative test.
(3)
In several of the tests performed using this procedure a noncenservative leak rate of zero was recorded when the instrumentation used was not sufficiently sensitive to justify this value.
The minimum sensitivity of the pressure decay test can be calculated using one-half of the smallest division on the pressure gauge, the test volume and the test time.
This minimum sensitivity concept is recognized by the licensee in his " Local Leak Rate Tests" procedure.
d.
Main Steam Isolation Valve (MSIV) Leak Testing The following item refers to the " Main Steam Isolation Valve Leak Rate Test" procedure and is designated unresolved item (219/79-08-03).
(1) The method used to leak test the irboard MSIV's may not measure the total leakage of the valves.
They are tested by pressurizing the reactor vessel and measuring the leakage from a flow meter attached to a test connection between the inboard and outboard MSIV.
This method uses the outboard MSIV as a leakage boundary and relies on it being leaktight.
Several test results were accepted with zero leakage recorded for an inboard valve and 7.5 to 8.6 SCFH (standard cubic feet per hour) recorded for the corresponding outboard valve.
There is no assurance that all leakage past the inboard MSIV was measured by the flowmeter and that none leaked past the outboard valve.
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5.
Document Retention It was noted in review of the above procedures that entire copies of completed surveillance procedures are not being retained.
In an apparent effort to reduce the bulk of stored documents, pages on which there is no writing other than the original printed procedure are being thrown away.
There was no documentation available describing this practice and it is apparently being conducted without the knowledge of the technical staff or the plant management.
This area will receive further review in a subsequent inspection. (219/79-08-04)
6.
Unresolved Items Items about which more infonnation is required to determine acceptability are considered unresolved.
Paragraphs 4b, c and d of this report contain unresolved items.
7.
Exit Interview At the inspection's end the inspectors held a meeting (see Detail 1 for attendees) to discuss the inspection scope and findings.
The unresolved items were identified.
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TABLE 1 November 1978 - CILRT Conducted at Pt (20 PSIG) at Oyster Creek Nuclear Generating Station Item Acceptance Criteria Reported Results Inspector's Findings 1.
Containment Leak Rate
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0.4265 0.433 (Mass Point)
2.
Upper 95% Confidence Level
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0.4298 0.437 on Leak Rate (ftss Point)
3.
Local Leak Rate Test
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0.0047 0.005 Additions (Type B&C)
4.
Containment Leak Rate at 0.567 0.4894 (Note 2)
0.497 (Note 2)
Upper 95% Confidence Level (0.75 Lt)
Plus Required Additions 5.
Type B&C Leak Rate Total 0.6 La 0.116 La 0.115 La 6.
Supplemental Verification 0.25 Lt 0.104 Lt 0.107 Lt Test Difference No tes:
1.
All data reported in weight percent per day unless otherwise indicated.
2.
Includes addition of 0.055 %/ day dee to water level increases inside containment.
Un ces CD tw L.~.
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