IR 05000327/1978043
| ML19269D089 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/10/1979 |
| From: | Dance H, Donat T, Yeomans B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19269D088 | List: |
| References | |
| 50-327-78-43, NUDOCS 7902270308 | |
| Download: ML19269D089 (10) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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Report No.: 50-327/78 43 Docket No.: 50-327 License No.: CPPR-72 Licensee: Tennessee Valley Authority 830 Power Building Chattanooga, Tennessee 37401 Facility Name:
Sequoyah Unit 1 Inspection at:
Sequoyah Site, Daisy, Tennessee Inspection conducted:
December 13-15, 1978 Inspectors:
R. H. Wessman A. H. Johnson Accompanying Personnel:
B. J. Yeomans T. J. Donat (December 15, 1978)
Approved by:
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H. C. Dance, Chief Date Reactor Proj ects Section No.1 Reactor Operations and Nuclear Support Branch Inspection Summary Inspection on December 13-15, 1978:
(Report No. 50-327/78-43)
Areas Inspected: Routine, unannounced inspection of preoperational test program administrative controls; observation of test activities in progress; review of containment leak rate test planning; and facility tour. The inspection involved 72 inspector-hours onsite by four NRC inspectors.
Results: Of the four areas inspected, no items of noncompliance or deviations were identified.
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RII Rpt. No. 50-327/78-43 I-1 DETAILS I Prepared by:Dh i/tb r)
R. H. Wessman, Reactor Inspector
' Datd Reactor Projects Section No. 1 Reactor Operations and Nuclear Support Branch Dates of Inspection: December 13-15, 1978 Reviewed by: hC JW i5'f1j
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H. C. Dance, Chief Date Reactor Projects Section No. 1 Reactor Operations r.nd Nuclear
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Support Branch 1.
Persons Contacted Tennessee Valley Authority (TVA)
- J. Balletine, Plant Superintendent
- W. Andrews, Plant QA Staff Supervisor
- E. Condon, Preoperational Test Section Supervisor S. Wilburn, Mechanical Engineer P. Garrett, Mechanical Engineer K. Hurt, Electrical Engineer M. Halley, Nuclear Engineer W. Kinsey, Assistant Power Plant Results Supervisor The inspector also interviewed five other licensee employees during the course of the inspection. They included QA Section, operations, and preoperational test section personnel.
- Denotes those present at the Exit Interview.
i 2.
Licensee Action on Previous Inspection Findings (0 pen) Unresolved Item (73-31-01): Test conduct ade nistrative discre-
pancies relating to W6.lB, SIS Accumulator Blowdown lests. The inspector determined that the applicant has implemented training for preoperational test section personnel. Test changes have been signed by the Test Section Supervisor but not by the Plant Superintendent. The Temporary Conditions Log did not fully reflect current plant conditions. Deficiency Notices reflected current status as tc test deficiency items and resolutions thereto. Test equipment information required by Section 4 of the test procedure was complete for equipment used in September testing. However, test equipment calii.ation data had not been recorded for equipment used in December testing. In that various administrative. discrepancies remain unresolved, this item remains open.
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RII Rpt. No. 50-327/78-43 I-2 (0 pen) Followup rtem (78-31-02): Modification to the SIS Accumulators.
The inspector rev; ewed the executed Work Plan No. 1308 for this modifi-cation. Until ret. sting and analysis of test data is complete this item remains open. The inspector has specifically requested this test data (and analysis thereof) for review.
(Closed) Followup Item (78-28-01): Diesel Generator Load Sequence Times.
The inspector had identified a possible discrepancy between the acceptance criteria of Preoperational Test 13 B(1) Onsite Power Distribution System (Diesel Load Sequencing), and FSAR Tatie 8.3-3.
The applicant had revised FSAR Table 8.3-3 in Amendment 56, indicacing that these values are nominal values only, closing this item.
(Closed) Followup Item (78-11-01): QA documentation discrepancies iden-tified in tentative transfer packages. These discrepancies, as well as other discrepancies listed in the tentative transfer package, are tracked via a computerized system maintained by construction. Review of workplans generated in the November - December 1978 period revealed evidence of con-struction effort to rectify discrepancies in tentative transfer packages.
(0 pen) Followup Item (78-11-02): Workplans should assure preoperational test results are not invalidated. See paragraph 5 for discussion of this item.
(0 pen) Unresolved Item (78-21-01): Westinghouse review of test procedures.
The applicant has established a program for obtaining Westinghouse review of pre perational test procedures pursuant to FSAR Paragraph 14.1.1.1.1.
Although Section 14.1.1.2 of the FSAR (Startup Test Program) is silent con-cerning Westinghouse reviews of startup test procedures in the NSSS - scope, Section 14.2.1.2 of the FSAR (Assisting Organizations) states that Westing-house "... will review test instructions applicable to the Nuclear Steam Supply System.. " In that a program for startup test instruction review has not been established, this item remains open.
(0 pen) Followup Item (Report 78-21 paragraph 5.c): Test Record Drawings not maintained in completed test data package. The applicant is initiating administrative controls to assure test record drawings (or microfilm copies thereof) are transmitted to the plant master file with the test data package.
This item remains open pending implementation and inspection of this program.
(Closed) Followup Item (78-11-03): Ventilation Systems Tests TVA 9A and 9C discrepancies. Acceptance criteria for these tests have been clarified.
A correction to T.I.9 (filter testing procedure referenced by TVA 9A and 9C) will clarify data calculations and acceptance criteria. Extrapolation
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RII Rpt. No. 50-327/78-43 I-3 of test data to allow evaluation of test data to allow evaluation of
" worst case" syst.-m capability is planned by the Division of Engineering Design (ENDES). Also, ENDES will participate in the test conduct for these tests.
3.
Unresolved Items None.
4.
Exit Interview The inspector met with Mr. J. M. Ballentine and members of his staff (as denoted in paragraph 1) on December 15, 1978. The inspector summarized the inspection findings. Also discussed was the applicant's program for workplan administration. The applicant agreed to develop additional con-trols in this area by January 15, 1979 (Paragraph 5).
5.
Use of Workplans (78-11-02)
The inspector and the applicant discussed the administrative controls relating to workplans and preoperational testing. The following items were identified as requiring resolution prior to closing outstanding item 78-11-02:
a.
A system does not exist to facilitate the tracking of outstanding workplans af fecting preoperational test conduct or the status of previously tested systems.
b.
Responsibility for conducting the system retest has not been defined.
Depending on the system status it may be the preoperational test section or P-Prod operations personnel.
Workplans which specify "preoperational test af fected" are silent as c.
to the test (or test steps) that require conduct to requalify the system. If a new (previously unwritten) test is required, such a test is not included with the workplan.
The applicant committed to develop and implement administrative controls relating to these items by January 15, 1979.
The inspector reviewed approximately 50 workplans issued during the November - December 1978 period.
Other than the comments identified above, inspector questions relating to these workplans were resolve.
RII Rpt. No. 50-327/78-43 I-4 6.
Conduct of W6.2, Upper Head Injection Test The inspectors observed portions of the conduct of W6.2, Upper Head Injection Test, and reviewed test documentation. Testing was determined to be conducted within the administrative controls of SQA 14, Sequoyah Nuclear Plant Preoperational Test Program, and the W6.2 test procedure.
In that test results data was not fully analyzed, review of test data and the completed test package will be conducted during a subsequent inspectica (78-43-01).
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RII Rpt. No. 50-327/78-43 II-1
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DETAILS II Prepared by:
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A.H.Joyngon,ReactorIpsp/ctor te Nuclear upport Section da. 1
Reactor Operations and Nuclear Support Branch
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Dates of Inspection: December 13-15, 1978 Reviewed by:
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[7e R. 5. Martin, Chief Nuclear Support Section No. 1 Reactor Operations and Nuclear
Support Branch u
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Persons Contacted Tennessee Valley Authority (TVA)
- J. Balletine, Plant Superintendent
- W. Andrews, Plant QA Staff Supervisor
- E. Condon, Preoperational Test Section Supervisor
- C. Cantrell, Assistant Plant Superintendent
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- K. Clark, Test Director, Outage L. Farmer, Test Director, Outage
- H. Fischer, Supervisor, Mechanical Engineering, Construction
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T. White, Engineer, Outage The inspector also interviewed seven other licensee employees during the course of the inspection. They included operations, conctruction and preoperation test section personnel.
2.
Licensee Action on Previous Inspection Findings
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Not Applicable to this Detail.
3.
Unresolved Items No new unresolved items this report period.
4.
Exit Interview The inspection scope and findings were summarized on December 15, 1978 with those persons indicated by an asterisk in Paragraph I above.
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this meeting the inspector discussed the areas inspected and sumarized this preliminary inspection as discussed in this report. In the areas inspected the licensee was informed that no apparent items of noncom-pliance or deviations were identified.
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RII Rpt. No. 50-327/78-43 II-2 5.
Containment Integrated Leak Rate Test (CILRT) and Local Leak Rate Testing (LLRT)
a.
The following items were discussed:
(1) For those systems.nat have salves tested in the reverse direction of containment pressure, the licensee must provide a documented evaluation to demonstrate that the results will provide equivalent or more conservative test results.
(2) The liceusee agreed to ascertair. that all commitments made to the NRC in the Question and Response Section of the FSAR are met by the Leak Rate Test Procedures. Responses to NRC questions address both local and integrated leakage rate testing.
(3) The licensee was made aware that the NRC requires that Type A Test (CILRT) instrument calibrations and certification to be traceable to National Bureau of Standards (NBS) or other recognized standards.
This documentation will be made available to the inspector prior to performance of the test.
(4)
The inspector received approved copies and changes to the local leak rate test (LLRT) procedures while on site.
The licensee stated that within three weeks the containment integrated leak rate test (CILRT) procedure would be approved and be available to the inspector ior review.
(5) The licensee was made aware that the isolation and venting of the pressurization (air compressors) source is required upon reaching test pressure and start of official CILRT test data collection.
(6) The licensee stated that some form of tagging would be used for valve line up of systems for the CILRT test.
(7) The Licensee stated that they would use an event log during the CILRT test.
(8) The licensee stated that local leak rate testing would be perf ormed af ter completion of the CILRT, on penetrations used to perform the CILRT test (i.e., penetration used to pressurize the containment).
(9) The licensee stated that, to maintain administrative controls during the CILRT, only one copy of the CILRT test procedure would be used and that this copy would be located at the CILRT Test Station.
(10) The licensee stated that the reactor coolant system would be vented on the reactor head during the CILRT test.
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RII Rpt. No. 50-327/78-43 II-3 (11) The licensee stated that standard practice SQA14 would be used to make changes to the approved leak rate test procedures.
(12) The inspector discussed problem areas in testing (CILRT) a containment with an ice condenser. The inspector noted that the licensee was e' ready aware of these problem areas because of their discussions with a licensee having a similar type containment.
b.
The inspector informed the licensee that the following NRC Staff positions would be used while performing the inspection of the licensee's leak rate test program:
1.
Venting and Draining Position The reactor vessel, those systems that are part to the reactor coolant pressure boundary and could provide direct communication with the containment atmosphere under post-accident conditions, and those systems which are postulated to rupture (i.e., not designed for post-accident function) shall be vented and drained to the extent necessary to assure exposure of the containment isolation valves (as defined in Section II-H) to the containment air test pressure, such that they will be subjected to the simulated accident differential pressure.
If the venting and draining of any system potert' lly jeopardizes the maintenance of a safe shutdown condit; i, then those systems shall not be vented and drained; aowever, in this event, the local leakage rates (Type C) for the isolation valves in these systems shall be added to the upper 95%
confidence limit of the CILRT before determining the accept-ability of the test.
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CILRT Correction for Local Leakage Position If, during the performance of a Type A test, :dentifiable local leakage occurs to the extent that it could cause failure of the Type A test, e.g.,
through penetrations or isolation valves, the leak may be isolated and the Type A test continued until completion.
A containment penetration which is isolated during a Type A test must have a design which will permit local leak testing of all potential leakage paths through the penetrations.
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RII Rpt. No. 50-327/78-43 II-4 Local leakage rates measured before and af ter repair must be reported, and the sum of the post-repair leakage rate and the CILRT upper 95% confidence limit must meet the Appendix J allowable leakage rate (0.75 La).
However, the difference in Type B and C test results before and after the repair of local leaks may not be deducted from the Type A test results in order to achieve an acceptable containment integrated leak rate.
Type B and C leak rate testing and repair prior to containment CILRT is also considered to be acceptable.
3.
Inclusion of Instrument Errors Position The uncertainty in the measured leakage rate shall be estab-lished by calculating the 95% upper confidence limit (UCL) of the least squares fit of the leakage rate data. The test results will be considered acceptable, if the 95% UCL is less than or equal to 75% La (75% LT for reduced pressure tests).
The accuracy of the measurement of the Type A leakage rate will be verified by the supplemental verification test. The measured difference between the supplemental verification test leakage rate and the Type A test leakage rate must be within 0.25 (0.25 Lt for the reduced pressure tests).
6.
Assurance of Meeting the Guidelines of 10 CFR 100 In reference to hTC letter of August 8, 1975 (Docket Nos. 50-327 and 50-328) from Mr. Karl Kniel, Chief, Light Water Reactors Branch 2-2, Division of Reactor Licensing to Mr. James E. Watson, Manager of Power, TVA, which required additional information for inclusion in the Safety Evaluation. Item 4 of this letter required the following:
4 Item 4 In order to reduce potential offsite doses from the design basis accident and assure meeting the guidelines of 10 CFR 100, we will require that:
a.
The primary containment leak rate be limited to 0.207; per day, b.
The allowable bypass leakage be limited to 2% of the primary containment leakage.
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c.
Leak paths that terminate in the auxiliary building be limited to 10% of primary containment leakage.
Recent Staff Technical Position CSB 6-3 " Determination of Bypass Leakage Paths in Dual Containment Plants" is attached for your information and use. We will require that all potential bypass leak paths be identified in accordance with the require-mednts of CSB6-3, and a surveillance requirement established.
Shorld your detailed analysis of bypass leakage in accordance wi'. the branch position indicate that this is required, it may be appropriate to further modify the integrated leak rate. "
This item will be held as an open item pending final dis-position by documented correspondence and/or final issuance of technical specifications. (50-327/78-43-01 and 50-328/78-43-01)
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