IR 05000312/1978002
| ML19309A157 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/23/1978 |
| From: | Canter H, Crews J, Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML19309A152 | List: |
| References | |
| 50-312-78-02, 50-312-78-2, NUDOCS 8003260871 | |
| Download: ML19309A157 (6) | |
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U. S. NUCLEAR REGULATORY C0ffiISSION OFFICE OF INSPECTION AND ENFORCE!4ENT
REGION V
Report No.
50-312/78-02
- Docket No.
50-312 License No.
DPR - 54 Safeguards Group Licensee:
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Sacramento Municipal Utility District j
P. O. Box'15830 Sacramento, California 95813 Facility Name:
Rancho Seco
Inspection at:
Rancho Seco Inspection Conducted:
January 30-31 and February 2-3, 1978 Inspectors:
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<: %[j ohnson, Reactor Inspector 1/nj,,
P. H Date Signeo
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H.dC nter, Reactor Inspector Date Signed Y
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hG. W h
n, Reactor Inspector Intern Date Signed Approved By:
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J. L.
fews(Thief,ReactorOperationsand 7 ate 41gned
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Nuclear Support Branch
Sumary:
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Inspection on January 30-31 and February 2-3,1978 (Report No. 50-312/78-02)
i Areas Inspected: Routine unannounced inspection of facility operation, organization and administration,10 CFR 21 reporting, reported licensee
events, Integrated Leak ' Rate Test report, and licensee action on previous I
inspection findings. The inspection involved 48 inspector-hours on-sita
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by three NRC inspectors.
Results:
No noncompliance items or deviations were identified.
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8003 260 h 7/
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IE:V Form 219 (2)
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i DETAILS 1.
Persons Contacted
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R. Rodriguez, Plant Superintendent / Manager, Nuclear Operations
- R. Columbo, Technical Assistant
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- D. Cass, Maintenance Supervisor N. Brock, Instrument Control Supervisor
- M. Carter, Shift Supervisor
- G. Coward, Plant Mechanical. Engineer
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D. Whitney, Plant Nuclear Engineer i
- J. Sullivan, Quality Arsurance Engineer A. Locy, Electrical Maintenance Foreman The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineer-ing staff, technicians, reactor operators, and maintenance personnel.
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- Denotes those present at the exit interview.
2.
Licensee Action on Previous I.;scection Findings I
(Closed) Inspection of Limitorque Valve Operators (50-312/77-06).
A previous inspection report discussed the licensee's plans to inspect valve operators similar to the BWST isolation valve which had previously
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i failed to operate, as described in LER 77-12. A licensee representa-tive stated that these inspections had been completed with no other
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inoperable valves observed, although a similar condition of wear, in varying degrees, had been found in some of. the other valve opera-tors. As stated in LER 77-12, the problem affects the ability of the motor operator to reengage following manual operation. A follow-up report submitted by the licensee on January 26, 1978, described -
the conditions observed and stated that the valves concerned would be cycled in automatic following manual operation to ensure proper reengagement and alignment for. automatic operation. The licensee was observed to have issued an operating order to this effect.
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(Closed) Minimum sensitivity of LLRT's (50-312/77-06). The use of
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minimum instrument sensitivities in accounting for local leak rate test (LLRT) measurements rather than logging penetration leakage as-0SCFMwasreviewedbytheliegnsee.
The values in question were on the order of a factor of 10 below related limits and therefore were not significant. Also, inaccuracies associated with assigning a minimum sensitivity numb?r for the LLRT measurements created other accuracy errors.
Therefore, the licensee will continue the past logging practices in the area of local leak rate testing.
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-W (0 pen) Test, vent and drain connections (50-312/77-06). A licensee representative stated that a signature space will be added to the LLRT data sheet to document the cognizant engineer's review of test results. The cognizant engineer's signature on the local leak rate test data sheet (SP 205.02) will indicate that all appropriate
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barriers to containment leakage during a postulated loss of coolant accident for the tested penetration have been tested and/or verified in the proper position. This includes test, vent and drain lines coming off lines penetrating the containment, between the containment isolation valves.
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3.
Organization and Administration Facility organization and administration were examined during the inspection.
Included were consideration of personnel qualifications, shift crew composition, and membership and qualifications of the Plant Review Comittee and Management Safety Review Committee.
A change in facility organizational structure was noted to be pending. This was described in a letter to the NRC dated January 10, 1978, and was still under discussion between the licensee and NRR.
No noncompliance items or deviations were identified.
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4.
Plant Ooerations
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A review of plant operations was conducted, including examination of selected logs and records, operating orders, jumper (bypass) log en-tries, and " reportable occurrence" reports.
During a tour of the plant, observations were made regarding the status of alarms, monitoring instruments, seismic re:traints, valve positions, piping system con-ditions, houseceeping, control room manning, and other plant conditions.
The status of the plant was discussed with licensed operators on duty.
No noncompliance items or deviations were identified.
5.
10 CFR Part 21 Reouirements
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The inspector conducted a review of the licensee's actions to implement 10 CFR Part 21. This included examination of the licensee's imple-menting procedure (QCI No.10) and of documents posted in accordance with Section 21.6 of Part 21. Review of procurement documents dated s'ubsequent to January 6,1978, showed none subject to 10 CFR 21. Only one condition had been considered by the licensee, and this was reported pursuant to Part 21 (letter, SMUD to Region V, dated Dec-ember 30,1977). The following comments were presented to the licensee following the review.
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O a.
The licensee's proeddure (QCI No.10) discussed the review and reporting of defects, but did not cover the reporting of failures to comply with NRC requirements (when related to a substantial safetyhazard). The inspector stated that this should be in-cluded.
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b.
The procedure provided for QA review of certain documents to identify deviations which should be reviewed for possible report-ing under 10 CFR 21. The procedure should also provide for possible initiation of a Part 21 report based upon deviations or noncompliance identified by other groups or individuals (this was included in the notice posted pursuant to Section 21.6).
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c.
The term " defect" was used in place of " deviation" four times in the QCI 10 description of the evaluation process (a " deviation" becomes a " defect" only after the evaluation determines it to be one).
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The inspector observed that certain information had been posted pursuant to Section 21.6, but noted that the manner in which it was posted might not assure permanence of posting.
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According to licensee representatives, the Management Safety Review Committee (MSRC) determined that posting pursuant to O
s Section 21.6 would be done only at three locations at the plant site. The inspector observed that activities subject to Part 21 are also conducted at the district offices in Sacramento, most
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notably procurement, design review, core management and quality assurance activities. A representative of the MSRC stated during a subsequent phone conversation that posting in accord-ance with Section 21.6 would also be provided at the district offices.
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l On December 30, 1977, the licensee submitted a. written Part 21 report to the regional office regarding an error in the computer software used for calculating DNBR valves.
Based upon review of the licensee's
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report and a related report fro'm Babcock and Wilcox, the Office of i
Inspection and Enforcement detemined that other Babcock and Wilcox facilities were not affected.
Review during the current inspection of the condition reported by the licensee established that corrective actions had been taken at Rancho Seco as described in the. licensee's December 30 letter.
No noncompliance items or deviations were identified.
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Review of ILRT Report The inspector verified that the first Type A test report - Reactor Containment Building Integrated Leak Rate test, dated January,1978-
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includes the information required to be reported by Appendix J to
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10 CFR 50. The test results and/or supporting information discussed in the report appeared to be consistent with design predictions and.
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I performance specifications.
The reported mass point leak rate at the 95% upper confidence limit was 0.026 weight-per-cent-per-day.
(acceptance criteria is 0.075 weight-per-cent-per-day.)
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The inspector noticed that penetration 62, the penetration that was
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used to pressurize and depressurize the containment, was not local leak rate tested after use.
The licensee had no requirement to test this penetration, but did commit during the exit interview to run a local leak rate test on P62 by March 1,1978. This item will be
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reviewed by the licensee for inclusion in the test procedure for the next Type A test.
No noncompliance items or deviations were identified.
7.
Licensee Event Followup O
The circumstances and corrective actions described-in Licensee Event Reports submitted since the previous inspection were verified. The details of each event had been reported to facility management and
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reviewed by the Plant Review Committee. Licensee Event Reports re-viewed were as follows:
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LER No. 77-20 - Failure of high pressure injection valve to open.
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LER No. 77-21 - Computer software error.
Discussions and log (review showed that the breaker for valve SFV-23812 had been inspected as stated in LER 77-20), with no obvious problems noted. During subsequent discussiens.by phone, licensee representa-tives stated that further review and discussions with those involved had led to the conclusion that a loose fuse holder had been the cause.
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l The licensee also stated that a followup would be submitted.
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No noncompliance items or deviations were identified.
8.
Independent Inspection Effort The inspector verified that it is not possible to disable the SFAS
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i feature (s) associated with reactor building pressure by capping the
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-5-O sensing lines or closing isolation valves inside the containment.
Calibration records were reviewed to verify that the calibration program on the pressure transmitters and indicators was up to date.
No noncompliance items or deviations were identified.
9.
Exit Interview
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The inspectors conducted an exit interview with licensee representatives (denoted in paragraph 1) at the conclusion of the inspection on January 3.
The scope and findings of the inspection were discussed.
Licensee representatives.made the following remarks in response to
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certain of the items discussed by the inspectors:
Stated that a signature space would be added to the LLRT data sheet to document test completion and the cognizant engineer's review of test results.
(Paragraph 2)
Stated that procedure QCI No.10 would be revised to incorporate the inspector's coments.
(Paragraph 5)
Subsequent to the inspection, licensee representatives stated during telephone discussions that posting in accordance with 10 CFR 21.6 would be provided at the District Office (Paragraph 5).
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