IR 05000220/1983013
| ML20024B713 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/15/1983 |
| From: | Bettenhausen L, Rekito W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20024B707 | List: |
| References | |
| 50-220-83-13, IEB-78-09, IEB-78-9, NUDOCS 8307110266 | |
| Download: ML20024B713 (10) | |
Text
50-220/821213 50-220/821215 50-220/821218 U. S. NUCLEAR REGULATORY COMMISSION Region I Report No. 50-220/83-13 Docket No. 50-220 License No. DPR-63 Priority
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Category C
Licensee:
Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility Name:
Nine Mile Point Unit 1 Inspection At:
Scriba, New York Inspection Conducted:
May 16-19 and 23-25, 1983 Inspectors:
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W. A. Rekito, Lead Reactor Engineer date sitned date signed h@
Approved By:
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L. H. Betfenhausen, Chief, Test Programs date signed Section, DETP
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Inspection Summary:
Inspection on May 16-19 and 23-26, 1983 (Report No. 50-220/83-13)
Areas Inspected: Routine, unannounced inspection of the containment leakage testing program including witness of the system Integrated Leakage Rate Test and local penetration testing; Licensee Event Reports; followup to IE Bulletin and previous inspection findings; and general tours of the facility. The
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inspection involved 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> onsite by one region based NRC inspector.
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Results: No violations were identified.
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i 8307110266 830622 PDR ADOCK 05000220
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DETAILS 1.
Persons Contacted The technical and supervisory personnel listed below were contacted:
- J. Aldrich, Operations Supervisor C. Coniber, Quality Assurance Engineer
- K. Dahlberg, Maintenance Superintendent
- W. Drews, Technical Superintendent R. Nield, Technical Services Engineer
- R. Pasternak, Licensing Engineer T. Perkins, General Superintendent
"T. Roman, Station Superintendent K. Shea, Quality Assurance Group Leader
- M.
Silliman, Administrative Assistant
- J. Spadafore, Superintendent of Technical Services P. Sweeney, I&C Foreman B. Taylor, I&C Supervisor L. Young, Consulting Engineer, Bechtel NRC Personnel
- S. Hudson, Senior Resident Inspector The inspector also talked with and interviewed other licensee personnel during the inspection. They included members of the operating, training, technical services, and maintenance staffs.
- denotes those present at the exit interview on May 25, 1983.
2.
Followup on Previous Inspection Findings (0 pen) Unresolved Item (220/75-21-01): This item relates to the licensee's compliance with 10 CFR 50, Appendix J and identified containment isolation valves for several systems which were not being leakage rate tested.
The licensee documented these areas in correspondence to the NRC and later completed several modifications which permit testing many of these valves as required by Appendix J.
The licensee explained his intention to reconcile recognized differences between the containment isolation system leakage rate testing program described in the FSAR and Technical Speci-fications and make a new updated and revised submittal to the NRC early during the next operating cycle. This item will remain open pending receipt and NRC review of the licensee's revised submittal for compliance with Appendix J.
(Closed) Noncompliance Item (220/75-21-06): This item relates to the fact that the feedwater system containment isolation stop-check valves were not leak rate tested as required by 10 CFR 50 Appendix J.
The licensee's written response to this item states that the subject check
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i valves in the Feedwater System were not relied upon to perform a contain-ment isolation function.
Subsequent to that response the licensee re-evaluated this position and revised his Containment Isolation Valve local leakage rate testing program to include the subject feedwater system stop-check valves. The inspector verified that these valves were tested during the 1981 refueling outage and were scheduled to be tested prior to restart from the current outage.
This item is closed.
(Closed) Unresolved Item (220/77-01-03): This item relates to reverse direction leakage rate testing of certain CIV(s).
The licensee evaluated all valves being tested in this manner and determined that in each case the results were equivalent or more conservative than if the valves were tested with differential pressure applied in the correct containment outleakage direction.
In the case of several valves, the licensee has completed modifications and/or revised the test method to conduct tests in the containment outleakage direction. The inspector reviewed licensee memoranda dated July 16, 1978 and July 22, 1982 which document this evaluation and further planned improvements of the containment leakage
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testing program. This matter is resolved.
(Closed) Unresolved Item (220/81-14-01): On April 13, 1983, SORC approved Revision 8 to Procedure NI-ISP-25.2 which specified that the test pressure for leakage rate testing all containment isolation valves would be 35.5 0.5 psig.
The inspector reviewed records of tests completed prior to April 13, 1981 and verified that each test was conducted at a pressure greater than 35 psig. The inspector further interviewed personnel responsible for leakage rate testing and each was aware of the minimum test pressure of 35 psig required by Technical Specifications and Appendix J.
This
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matter is resolved.
(0 pen) Unresolved Item (220/81-14-03): This item relates to the plant history of containment leakage testing problems.
It identified the need for an engineering evaluation to assess the adequacy of the containment design and testing program to assure leaktight integrity throughout the operating cycle. The licensee conducted an evaluation of previous contain-ment leakage problems and N.s upgraded the containment system design by modifying or replacing c'. 0 t ontainment isolation valves (CIVs). The licensee explained t M 34a ticnal improvements to the containment system design were planne' 2 > tF he was preparing a revised submittal to the
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NRC describing the;. ae;<.e af conformance with 10 CFR 50 Appendix J.
l This revised submittal is also intended to reconcile recognized differences between the containment isolation system leakage rate testing program described in the FSAR and Technical Specification. The inspector reviewed a licensee memorandum dated July 22, 1982 which documents part of the engineering evaluation conducted and examined several of the CIV modifica-tions completed during the recent outage. The licensee's actions completed and planned appear to be appropriate and acce m ble.
However, this item will remain open pending NRC receipt and further review of the licensee's revised submittal for compliance with Appendix.
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(Closed) IE Bulletin 78-09: BWR Drywell Leakage Paths Associated with Inadequate Drywell Closures. A previous review during NRC inspection 220/81-14 found that controls established for the drywell head installation met the bulletin guidelines except for assuring reproducible tightness.
The licensee subsequently revised his installation method to utilize a hydraulic stud tensioning device. The inspector reviewed Procedure MP 1.8, " Installation of the Containment Head", Revision 5 and observed the special tensioning tool in use on May 15, 1983.
Based on the controls established by the licensee and the lack of leakage problems experienced at drywell closures, this bulletin is closed.
However, the inspector noted that the current practices are different from those described in a letter dated July 28, 1978 submitted to the NRC in response to this bulletin.
The licensee stated that he would amend that letter by submitting an updated response reflecting their current practice.
3.
Containment Local Leakage Rate Testing (LLRT)
3.1 Documents Reviewed
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Procedure NI-ISP-25.2, Revision 7, Primary Containment Isolation - Leak Rate Tests
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Procedure NI-ISP-23.4, Revision 2, Local Leak Rate Test -
Summary
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Procedure NI-ISP-24.7, Revision 1, Core Spray System Check Valves - Leak Rate Test
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Procedure NI-ISP-25.5, Revision 1, Primary Containment Isolation - Leak Rate Tests
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Procedure NI-ISP-25.6, Revision 1, Primary Containment Isolation - Leak Rate Tests
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Procedure NI-ISP-25.7, Revision 0, Feedwater Isolation Check Valves - Leak Rate Tests
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Procedures ISP-IC-25.1 A thru G, Revision 0, Main Steam Isolation Valve Leak Rate Tests
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Licensee Event Reports 82-19 thru 23, Abnormal degradation of systems designed to contain radioactive material resulting from fissior process
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Records of LLRT activities conducted during the recent outage including test results, related repair and retest l
documentation
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Selected system drawings, Piping and Instrumentation Diagrams.
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3.2 Scope of Review The inspector reviewed the above documents to ascertain compliance with regulatory requirements of 10 CFR 50 Appendix J, Technical Specifications, and applicable industry standards. The inspector also witnessed selected LLRT activities and held discussions regarding the relationship between LLRT failure results, the subsequent leakage improvement, and evaluation of the "AS FOUND" leakage condition for the Containment Integrated Leakage Rate Test (CILRT) results.
Further details and inspection findings are described below.
3.3 Test Witnessing On May 18, 1983, the inspector witnessed LLR1'r of the drywell head seal and a newly installed Core Spray system ulve 93-71. The tests were conducted in accordance with an approved procedure and the results were acceptable. Additionally, the inspector observed that the test instrument was properly calibrated and the test personnel appeared to be knowledgeable of the procedure requirements and suitably qualified. These test activities were also witnessed by station Quality Assurance Representatives as part of their outage
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surveillance audit program.
In discussion with the inspector, the QA engineers explained that they routinely witness several containment LLRT(s) each outage. No unacceptable conditions were identified.
3.4 Test Results The inspector reviewed the LLRT Results Summary and discussed analysis of test results and the status of repairs and retests with the licensee. A problem was discovered with the Feedwater system check valves 31-01 and 31-02. The licensee's investigation revealed that the excessive leakage condition was caused by foreign material l
(welding slag) introduced into the valve's teflon seating surface j
during the previous outage when the adjacent gate isolation valve was l
replaced.
Several other CIV(s) had to be repaired because of exceeding leakage limits but none resulted in a more significant problem of unacceptable leakage past both redundant barriers. The inspector explained that the overall primary containment system leakage improve-ments resulting from these repairs prior to the CILRT are expected to be added to the results of the CILRT for determination of the "AS IS" Containment System leakage condition in accordance with 10 CFR 50 Appendix J.
The requirement was acknowledged by the licensee and his preliminary evaluation of LLRT results for this purpose does not indicate any problem with the success of the CILRT.
3.5 Licensee Event Reports (LER's)
By letter dated January 12, 1983, the licensee submitted five LERs (82-19 thru 23) identifying CIV(s) which were found to exceed their Technical Specification allowable leakage rate limit fo 12.9 SCFH at 22 psig.
The inspector verified by record review and personnel
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interview that:
the events were clearly and accurately described including determination of cause; review for generic implications was conducted; appropriate corrective actions had been completed; and reporting requirements of Technical Specifications met. No unacceptable conditions were identified.
4.
Primary Containment Integrated Leakage Rate Test (PCILRTJ 4.1 General On May 23-25, 1983, the licensee performed a PCILRT as required by TS 4.3.3.
The test was conducted at a test pressure of 22 psig in accordance with Procedure ISP-IC-23, Revision 13.
The inspector reviewed the test procedure, witnessed preparations and various portions of the test and verified qualifications of test personnel.
Further details and inspection findings are described below.
4.2 Procedure Review The inspector reviewed the PCILRT procedure along with documents listed in Paragraph 3.1 for technical adequacy and to ascertain compliance with requirements of Technical Specifications and 10 CFR 50, Appendix J.
The inspector noted that the procedura required a minimum test duration of eight hours.
This reduced duration test is acceptable to the NRC based on the procedure meeting the requirements of Bechtel Power Corporation's Topical Report BN-TOP-1, " Testing Criteria for ILRT of Primary Containment Structures for Nuclear Power Plants." The procedure referenced and was in general conformance with industry standard ANS/ ANSI 56.8-1981, Containment System Leakage Testing Requirements. No unacceptable conditions were identified.
4.3 Test Boundaries On a random sampling basis, the inspector reviewed the procedure valve lineups for approximately 25 percent of the piping penetrations.
This review was to ensure that systems were properly vented and drained to expose the CIVs to containment atmosphere and test differential pressure with no artificial leakage barriers.
No problems or unacceptable conditions were identified.
l 4.4 Instrumentation The inspector reviewed the calibration records for resistance temp-erature detectors, dewpoint instruments, precision pressure detectors and verification test flowmeters. Their calibrations prior to the PCILRT were found to meet applicable accuracy requirements and were traceable to the National Bureau of Standards. The inspector also verified that the instrument system specified in the test procedure I
satisfied the instrument selection guide of the Bechtel Topical l
Report BN-TOP-1 and ANSI /ANS 56.8.
The inspector observed operation l
of the automatic data collection system during conduct of the test.
j No unacceptable conditions were identified.
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4.5 Inspection Tours The inspector conducted inspection tours with licensee personnel both before and during the PCILRT. During these tours, test boundaries were surveyed for evidence of leakage, and on a sampling basis, selected valves were verified to be in the correct positions according to procedure requirements.
No unacceptable conditions were identified during these tours.
However, the licensee discovered a significant air leak at the mechanical seal of the Core Spray Topping Pump No. 121.
To more accurately simulate the post accident containment isolation system condition this pump was placed in service and the integrity of the mechanical seal was demonstrated when no appreciable water leakage was observed.
Following this demonstration an additional ring of packing was added to the affected mechanical seal and the test was resumed.
The above event was documented in the PCILRT events log and adequately controlled by the testing staff.
No unacceptable conditions were identified.
4.6 PCILRT Chronology Date Time Description May 23 Completed test preparations and containment inspection.
2120 Commenced pressurizing the containment.
Emptied three pressurized bottles of Argon gas into the drywell to aid in leak detection.
May 24 Conducted leak surveys.
0305 Completed pressurizing the containment at test pressure (P ) f 23.1 psig.
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Commenced data collection and continued leak surveys.
0500 Preliminary calculations indicate unacceptable leakage rate of approximately 1.9 percent per day.
0600 Discovered significant air leakage at the mechanical seal of Core Spray Topping Pump No. 121.
1130 Secured the CRD pumps and isolated the cleanup system.
1315 Conducted local leak rate test of shutdown cooling system valve 38-1.
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Date Time Description 1800 Core Spray system operated to demonstrate integrity of pump mechanical seals under normal operating pressure and water conditions.
Additional ring of packing added to mechanical
seal of Core Spray Topping Pump 121.
2230 Calculated leakage rate improved and appeared acceptable.
May 25 0715 Completed minimum 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> test period of data collection.
0720 Imposed a leak of 4.5 SCFM and started the leak rate verification test.
1230 Completed the verification test and commenced de-j pressurization of the containment.
4.7 PCILRT Results The licensee evaluated the test results for the minimum eight hour period from 2230 on May 24, to 0715 on May 25, 1983. The calculated leakage rate at the upper 95 percent confidence limit was 0.305 weight percent per day. The test acceptance criterion is 0.82 weight percent per day.
The inspector noted that the above calculated leakage rate did not include required corrections for certain penetrations in use or isolated during the test. A preliminary conservative estimate for the combined leakage correction is approximately 0.210 weight percent t
per day. Additionally, this value represents the Containment System
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"AS LEFT" overall leakage. An evaluation of any leakage improvement resulting from repairs performed prior to the PCILRT is needed to determine the Containment System "AS FOUND" overall leakage. The
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inspector stated that both values are expected to be reported in the Technical Summary Report. The licensee stated that this would be done and that a preliminary evaluation of the "AS FOUND" condition indicated system leakage well within the maximum allowable by Technical Specifications of 1.1 weight percent per day at a test pressure of 22 psig.
The inspector independently calculated several mass values and leakage rates using raw data from the test. The results were in agreement with and verified the accuracy of the licensee's L akage rate calculation r
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4.8 Quality Assurance Involvement As part of the QA Department surveillance audit plan, the PCILRT was monitored by several QA engineers. The inspector reviewed their prepared Surveillance Check List No.83-022.
He also held discussions regarding their observations and assessment of the PCILRT prc edural controls as well as the containment leakage testing program at?quacy.
No problems or unacceptable conditions were identified.
5.
Containment Continuous Leakage Monitoring Technical Specification surveillance requirement 4.3.3.g states that when the primary containment is inerted, the containment shall be con-tinuously monitored for gross leakage by review of the inerting system makeup requirements. The inspector inquired as to how this TS surveillance was being satisfied. The licensee explained that during previous operating cycles reliance was placed on the operating staff and management overview for nitrogen makeup and consumption records.
In addition, the licensee identified a TS Change Request to delete this requirement submitted to the NRC on January 11, 1979.
This request is still under review by NRR.
Subsequently the licensee added a data sheet (Fig. OP-9-10) to Revision 8 of Operating Procedure NI-0P-9, Nitrogen Inerting and H -0 Monitoring 2 2 Systems for Primary Containment. The licensee plans to use this data sheet to maintain an account of nitrogen makeup requirements during normal operations.
The inspector reviewed the operating procedure and noted that no specific reference to the TS requirement was made. Also, the procedure lacked guidance for evaluating this data and establishment of appropriate acceptance criteria. The licensee acknowledged the inspector's concern for a documented formal evaluation of the C
.a but stated that this would be done as part of the operating superintei 'ent's review of records until confidence was gained for data reliability and appropriate procedures revised or developed.
The inspector considered the licensee's planned action to be acceptable.
However, this matter is unresolved pending further NRC review of the licensee's actions to establish a more formal evaluation of available data to satisfy TS surveillance requirement 4.3.3.g.
(Unresolved Item 220/83-13-01)
6.
Facility Tours The inspector made several tours of the facility including the turbine building, reactor building, refueling floor, relay room and the control room.
During these tours, the inspector observed operations and activities in progress, implementation of security and radiological controls, general condition of safety related equipment, component tagging and system operations to support the PCILRT.
No unacceptable conditions were identified as a result of these observations.
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7.
Unresolved Items i
Items about which more information is required to determine acceptability are considered unresolved.
Paragraphs 2 and 5 contain unresolved items.
8.
Exit Interview The inspector met with licensee representatives (see Detail 1 for attendees)
On May 25, 1983. The inspector ',ummarized the scope and findings of the inspection at that time.
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