IR 05000282/1988004

Insp Repts 50-282/88-04 & 50-306-88-04 on 880214-0402. Violation Noted.Major Areas Inspected:Previous Insp Findings,Plant Operational Safety,Maint,Surveillances,Esf Sys,Ler Followup & Compliance Bulletin Followup

ML20151S934
Person / Time
Site:	Prairie Island
Issue date:	04/18/1988
From:	Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20151S931	List: IR 05000282/1988004 ML20151S930
References
50-282-88-04, 50-282-88-4, 50-306-88-04, 50-306-88-4, IEB-87-002, IEB-87-2, IEB-88-001, IEB-88-003, IEB-88-1, IEB-88-3, IEIN-87-041, IEIN-87-063, IEIN-87-41, IEIN-87-63, NUDOCS 8804280521
Download: ML20151S934 (10)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-282/88004(DRP); 50-306/88004(DRP)

Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Mall

• Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plant Inspection At: Prairie Island Site, Red Wing, Minnesota

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Inspection Conducted: February 14 through April 2, 1988 Inspectors: J. E. Hard M. M. Moser Approved By: Y//f/B Reactor Projects Section 2A Dat'e

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Inspection Summary Inspection on February 14 through April 2,1988 (Reports No. 50-282/88004(ORP);

No. 50-306/88004(0RP))

Areas Inspected: Routine unannounced inspection.by resident inspectors of

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previous inspection findings, plant operational safety, maintenance, surveillances, ESF systems, LER followup, compliance bulletin followup, information notice followup, closecut of instructions, meetings with corporate management, and meetings with public official Results: Of the ten areas inspected, one violation was identified; however, in accordance with 10 CFR 2, Appendix C, Section V.A., a Notice of Violation was not issued (failure to follow procedures - Paragraph 3).

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8804280521 880419 PDR ADOCK 05000282

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DETAILS Persons Contacted -

• • J. Howard, President and Chief Executive Officer

• • McCarthy, Chairman of the Board

• • Jensen, Senior Vice President, Power Supply

• • Gilberts, Senior Vice President Special Projects

• • Larson, Vice President, Nuclear Generation '

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• • L. Eliason, Ger.eral Manager, Nuclear Plants

• • G. Neils, General Manager Headquarters Nuclear Group

• • F. Tierney, General Manager Nuclear Engineering and Construction

• • D. Musolf, Manager Nuclear Support Services P. Kamman, Superintendent, Nuclear Operations Quality Assurance

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• E. Watzl, Plant Manager 00. Mendele, General Superintendent, Engineering and Radiation Protection

• Lindsey, Assistant to the Plant Manager

• Sellman, General Superintendent, Operations D. Schuelke, Superintendent, Radiation Protection G. Lenertz, General Superintendent, Maintenance K. Beadell, Superintendent, Technical Engineering M. Klee, Superintendert, Quality Engineering R. Conklin, Supervisor,' Security and Services D. Vincent, Project Manager, Nuclear Engineering and Construction J. Goldsmith, Superintendent, Nuclear Technical Services

• A. Hunstad, Staff Engineer S. Hiedeman, System Engineer T. Amundson, Superintendent Training A. Smith, General Superintendent, Planning and Services A. Vukmir, Site Services Representative, Westinghouse Electric Cor C. Gerstberger, Fueling Service Manager, Westinghouse Electric Cor D. DiIanni, License Project Manager, NRR The inspectors interviewed other licensee employees, including members of the technical and engineering staffs, shift supervisors, reactor and auxiliary operators, QA personnel, Shif t Technical Advisors, and Shift Manager * Denotes those present at the exit interview of April 4, 198 **Danotes NSP personnel who were visited by Region III personnel on February 19, 198 . Licensee Action On Previous Inspection Findings (92701)

(Closed) Vin 1ation (282/85012-01; 306/85009-01).

Failure to measure turbine driven auxiliary feedwater pump (TDAFP) speed and wait five minutes before obtaining pump test data. This violation was administrative 1y closed by the Division of Reactor Safety (DRS).

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(Closed) Open Item (282/85012-02; 306/85009-02).

Submittal of Containment Isolation Valve (CIV) leak rate testing relief to NR This open item was administratively closed by. OR (Closed) Violation (282/85012-03; 306/85009-03).

i Failure to treat vibration measurement instruments (i.e.,

IRD-306 metersi as measurement and test equipment for operability determination. This violation was administrative 1y closed by DRS.

l (Closed) Open Item (282/85012-04; 306/85009-04).

Inclusion of valve stroke method in procedure This open. item was administrative 1y closed by DR (Closed) Open Item.(282/F03078780; 306/F03078780).

NRC Position on Review Committee Quorum. Polling of Operations C/ nittee members by telephone is not permitted by current plant directive However, telephone conference calls which will permit interaction between members are permitted though infrequently used. These practices regarding the Operations Committee meet NRC requirement . Operational Safety Verification (71707)

l Unit I was base loaded at 100% power except for reductions for i

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surveillance testing and a forced outage from March 10 to March 14 for repair of a leaking seal weld on a reactor vessel head instrument tub .

Unit 2 was base loaded at 100% power except for reductions for l surveillance testin l The inspector observed control room operations, reviewed applicable logs, I conducted discussions with control room operators, and observed shift l turnovers. The inspector verified operability of selected emergency '

systems, reviewed equipment control records, and verified the proper return to service of affected components. Tours of the auxiliary building, turbine building and external areas of the plant were conducted to observe plant equipment conditions, including potential fire hazards, and to verify that maintenance work requests had been initiated for equipment in need of maintenanc On March 10, 1988, with Unit 1 at 100% power, sample results from the Unit 1 containment sump indicated a small amount of reactor coolant system leakage. Inventory measurements showed the rate of leakage to be very low, less than 0.1 gpm. A visual inspection inside the containment found that a reactor vessel head instrument column seal weld was leakin Unit 1 was taken off line at 2:06 p.m. and repairs were successfully completed on March 12, 198 ,.

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During Unit 1 startup preparations on March 13,~1988, reactor coolant system heat up was begun before the steps required in tha startup procedure were complete Specifically, the following: step in C1.2, Unit Startup Procedure was inadvertently omitted prior to starting heatup:

5.21 Lineup of containment spray and safety injection system When the omission was discovered by the Shift Supervisor, heatup was immediately stopped to complete these checks. At this time the average ,

reactor coolant temperature was about 198 degrees F., so the technical specification limit of 200 degrees F. was not exceeded. However, failure to follow the procedure is a violation of Technical Specifications Paragraph 6.5 (282/88004-01(DRP)). Corrective action was taken iamediately to correct the problem and this violation meets the tests of 10 CFR 2, Appendix C, Section V.A; consequently, no Notice of Violation will be issued, and this raatter is considered close . Maintenance Observation (62703)

Routine, preventive, and corrective maintenance activities (on safety-related systems and other balance of plant components) were observed / reviewed to ascertain that they were cenducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications. The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were ,

accomplished using approved procedures and were inspected as applicable, functional testing and/or calibratinns were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, radiological controls were implemented, and fire prevention controls were implemente Portions of the following maintenance activities were observed / reviewed during the inspection period:

Unit 1 Condenser Tube Cleaning ,

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l Unit 1 Fire Protection Panel Repair l Unit 2 #2R22 Shield Building Vent Gas Monitor Unit 1 Reactor Vessel Head Instrument Tube Seal Weld Repair Unit 1 Reactor Charging System Flow Control Valve Leakage Repair Emergency Cooling Backflush Line Leakage Repair No violations or deviations were identifie . Surveillance (61726)

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The inspector witnessed portions of surveillance testing of safety-related systems and component The inspection included verifying that the tests were scheduled and performed within Technical Specification requirements by observing that procedures were being followed by qualified operators, that Limiting Conditions for Operation (LCOs) were not violated, that system and equipment restoration was completed, and that test results were acceptable to test and Technical Specification requirement Portions of the following surveillances were observed / reviewed during the inspection period:

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SP 2093 Emergency Diesel Generator No. 2 Surveillance

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SP 2032 Unit 2 Reactor Safeguards Logic Test

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SP 2130 Unit 2 Containment Vacuum Breaker Test

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SP 1745 Diesel Generator No. 3 Surveillance (Non-Safeguar,d)

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SP 2005 Unit 2 Thermodynamic Heat Balance

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SP 1728 Siren Cancel Test

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SP 1088 Unit 1 Safety Injection (SI) Pump Surveillance

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SP 1112 Steam Exclusion Surveillance

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SP 2006 Nuclear Power Range Axial Offset Calibration

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SP 1116 Unit 1 Monthly Power Distribution Map No violations or deviations were identifie . ESF System Walkdown (71710)

The inspector performed a complete walkdown of the accessible portions of Unit 1 and Unit 2 safety injection systems. Observations included confirmation of selected portions of the licensee's procedures, checklists, plant drawings, verification of correct valve and power supply breaker positions to insure that plant equipment and instrumentation are properly aligned, and local system indication to insure proper operation within prescribed limit No violations or deviations were identifie . Licensee Event Reports and Part 21 Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications:

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(Closed) LER 282/87014-LL; Unit 1 power range hi flux low setpoint U

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exceeded the tech spec limi On May 28, 1987, Unit 1 was in its power escalation program following the Cycle 11-12 refueling. ' Calibration of the NIS power range channels is done based on calorimetric data taken during the startup physics testing program. At.an indicated power of 34.6%, calorimetric data showed that actual power was 45.7%. The effect of this inaccuracy is that the power range high flux low setpoint would have tripped the reactor at about 33% power; the Technical Specification setpoint given for this trip is 25%.

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The cause of this inaccuracy was a greater-than predicted change in radial leakage resulting from the change in core loading pattern from Cycle 11 to Cycle 12. A license amendment request'is being made to increase the required setpoint to make it consistent with the intermediate range setpoint. In addition, checks will be made of the power range channels at low power during startup testin . Compliance Bulletin Followup (25026)

(Closed) Compliance Bulletin No. 87-02 and Temporary Instruction 2500/26 (282/87-02-BB; 306/87-02-88): Compliance Bulletin No. 87-02, "Fastener Testing To Determine Conformance With Applicable Material, Specifications" dated November 6, 198 '

In a memo from C. E. Norelius, Director, Division of Reactor Projects, dated November 25, 1987, the Senior Resident Inspector was requested to review the actions taken by the licensee in response to the Compliance

Bulletin 87-02 and to actively participate in the licensee's fastener -

selection process in accordance with Temporary Instruction 2500/26. The i resident inspector reviewed the licensee's fastener procurement program, identified the fasteners of special interest, and was present during the selection of samples from warehouse supplies. A final review of selected  !

samples was made to ensure that as much of the selection criteria as l possible in TI 2500/26 were met prior to delivery to an independent -

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testing laborator .

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On January 26, 1988, the licensee issued a detailed report providing all 1 of the information that was requested in the compliance bulleti Results of the material and chemical tests of the selected safety related ,

and non-safety related fasteners found all but four non-safety related

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fasteners to be within specification The four non-safety related l fasteners in question were marginally out of specification ranges for ,

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hardness. However, being non-safety related and well within engineering safety factors and tolerances, no safety significance or equipment functional concerns exist.

It appears that, based upon testing of selected fasteners, the licensee has not identified any counterfeit or incorrectly marked-fasteners and

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that existing procurement procedures and administrative requirements are -

adequate to control the use of fasteners at Prairie Islan t

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i (Closed) Compliance Bulletin No. 88-01:. (282/88-01-88; 306-01-88): l

"Defects in Westinghouse Circuit 8reakers" dated February 5, 1988 In a letter dated March 16, 1988, the licensee indicated that the specific Westinghouse breaker in question (i.e., Westinghouse DS series)

is not used in Class IE service at Prairie Islan (Closed) NRC Bulletin Nc. B8-03:

Inadequate Latch Engagement in HFA Type Latching Relays Remanufactured by General Electric (GE) Company Review by the licensee of subject relays has determined that there are no GE HFA type latching relays installed in Class IE (safety related)

applications at Prairie Islan . Information Notice Followup (92701)

NRC Information Notice (IN) No. 87-41: Failures Of Certain Brown Boveri Electric Circuit Breakers In a memo dated from C. E. Norelius, Director, Division of Reactor Projects, dated October 27, 1987, the Prairie Island Senior Resident Inspector was requested to review the actions taken regarding IN 87-41 and whether the generic implications have been adequately considere The licensee completed a review of IN No. 87-41 and in addition to -

incorporating the close latch asti-shock spring (CLASS) modification

(85L858) on all subject 4.16 KV switchgear breakers has added a requirement to the preventive maintenance procedure to visually inspect the mounting bolt and closing' spring on the charging motor annually. The

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CLASS modification will be incorporated on all 4.16 kv switchgear -

breakers at the end of the Unit 1 refueling outage in September, 198 NRC Information Notice (IN) No. 87-63: Inadequate Net Positive Suction l Head in Low Pressure Safety Systems <

On February 18, 1988 while reviewing NRC Information Notice No. 87-63, '

"Inadequate Net Positive Suction Head in Low Pressure Safety Systems", a '

procedural inadequacy was discovered in the Prairie island Emergency Operating Procedures covering operation of the low head safety injection i

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pumps in the recirculation mode following a large break loss cl coolant acciden ,

The Emergency Operating Procedures were found to allow simultaneous use -

of low head upper plenum vessel injection, high head cold leg vessel injection, and containment spray while in the recirculation mode. It was determined during the investigation that the net positive suction head requirements for the RHR pumps, which serve as low head safety injection pumps, would not be met under these condition ,

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A Special Order was immediately issued providing guidance to the operators to assure adequate' suction head is available to the low head pumps under accident conditions. The Emergency Operating Procedures will be revised to fully address this concer A Westinghouse detailed flow requirement analysis for' Prairie Island is being prepare Changes to the Prairie Island Emergency Operating Procedures will be made based on the results of this detailed analysi Procedure changes will provide assurance that the RHR pumps in pos accident recirculation mode are always operated within their limitation ,

Revision of the Emergency Operating Procedures, which will include other desirable improvements, as well as necessary training of operators on the revised procedures, will.be completed by June 30, 198 . Closecut of Temporary Instruction (TI) (92703)

Primary Coolant System Pressure Isolation (Event V) Valves, TI 2515/84 Inspections required by this TI were completed with the following results:

06.01 Technical Specifications (TSs) were appropriately modified as required by the NRC Order to Northern States Power Co. dated April 20, 198 .02 Test procedure The testing method is judged to be acceptable. Direct volumetric means are used to measure rates of leakage through most of the check valves. Where direct flow measurement is not practical, rate of

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pressure buildup in known volumes or rate of level increase in tanks of known diameter are used to compute leak rates. Certain qualitative measurements such as in Step 20 of SP 2070,

"Qualitatively measure this flow by feeling the tubing leading to 251-20-20 and listening for flow.", are performed in preparation for actual leak rate measurements. In response to the inspector's queries on this matter, the licensee in the future will be making actual temperature measurements during these procedure step In the test procedures, more than one valve are tested simultaneousl However, acceptable leakage rates for any group of valves is limited to one gpm, the acceptable value for individual valves, Leakage rate limits for testing done at less than full pressure.are appropriately adjusted assuming leakage to be directly proportional to the pressure differential to the one-half powe Acceptance criteria in the test procedures are in accordance with !

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the T In the event of unacceptable leakage rate results, the procedure users are referred to the appropriate TS section to I determine corrective action neede )

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06.03 Review of records Records of Unit 1 testing in 1984 and Unit 2 testing in 1985 were reviewed as well as observance of a current test (see below). The records contain the major test data needed_to support the leak rate calculations and also contain the acceptance criteri Review of 1983 to 1988 test records showed that testing had been performed on the frequency required-by TS except that the valve 2SI-6-2 data from Unit 2 1984 refueling outage could not be located by the licensee. This missing point data comprises an Open Item, pending further search by the license (282/88004-02; 306/88004-01(DRP)). Valve leakage testing is done at the end of outages prior to reactor startu Measurements made at this time are in the condition the valves would be in at the start of unit operatio ' Leakage rate trending and analysis - Trend tables provided by the licensee were neither complete nor up to date. The inspector found it necessary to review individual completed surveillance documents in order to determine in many cases the actual leak rate values. This failure to maintain complete and orderly records and trend of leak rates, is similar to the situation noted in a previous inspection. See Inspection Report 50-282/84-08(ORS);

50-306/84-07(DRS), Section 3.b. During that inspection, the inspector was informed that the licensee agreed to track valve leak rates for trending purposes. At the time of the current inspection this commitment appeared not to have been effectively implemente This is an Open Item (282/88004-03; 306/88004-02(DRP)). i During the inspection no test data anomalies were noted which would have indicated improper or inaccurate testin No examples of Event V valves not meeting their acceptance j criteria were noted during this inspection. Licensee personnel ,

stated that there have been no known failures of these valves l to pass their test '

06.04 Performance of SP 2070, Reactor Coolant System Integrity Test was witnessed during the Unit 2 outage of Jan.-Feb. 1988. This surveillance includes a leak rate measurement of all except one of the Event V valve The inspector verified that test prerequisites were satisfied, that leakage rates were accurately determined, and that post-test activities and documentation were completed correctl . Meeting with Corporate Officials On February 19, 1988, NRC and licensee representatives met at the NSP Corporate Headquarters in Minneapolis to discuss matters related to the safety of operations at the Monticello and Prairie Island plant The ,

following subjects were discussed:

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NRC-initiated subjects Safety performance of nuclear plants Performance indicators SALP process Performance of Region III plants Observations on plant housekeeping Actions taken as result of Information Notice 87-13 revie Region III organizational change Licensee-initiated subjects Current NSP organization Licensee goals for improving performance Performance of SSFIs ,

Role of Corporate QA organization Getting feedback from NRC 12. Meetings with Public Officials (94600)

On March 10, 1938 NRC representatives including C. Paperiello, R. DeFayette, J. Hard, and R. Lickus met with local public officials in the Citizens Security Building meeting roo Officials present for the meeting were R. Kosac, Red Wing Fire Chief and A. Dolezal Hennigan, Minnesota Department of Health. J. Bohn also attended as an interested member of the public with a question from the Prairie Island Tribal ,

Council relative to restoration or replacement of Indian property in the event of a serious accident at the plan (This question was answered later in a letter to the Prairie Island Tribal Council from R. Lickus dated March 17, 1988). Two other members of the public were present along with three individuals who work at the plan The NRC representatives present reviewed the organization and charter of the ,

agency plus the roles of Region III individuals and of the resident inspector . IAEA Visit On February 15, 16 and 17, 1988,<two members of a team from the ,

International Atomic Energy Agency visited Prairie Island as part !

of a program by the agency to provide good practices information to third world countries interested in nuclear powe Prairie Island was one of eight nuclear plants chosen around the world by virtue of its good operating record of availability, low forced outage and trip rates, low personnel exposure, et During their visit, the team members met briefly with the resident inspectors and discussed the regulatory role of the NR . Exit (30703) j

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The inspectors met with the licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on April 4, 1988. The inspectors discussed the purpose and scope of the inspection and the findings. The inspectors also discussed the likely information content I of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any document / processes as proprietar I i

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