IR 05000155/1988009

Insp Rept 50-155/88-09 on 880420-0601.No Violations Noted. Major Areas Inspected:Site QA Staffing,Operational Safety Verification,Maint Observations of Work on Emergency Diesel Generator & NRC Bulletin 88-03

ML20150B509
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/29/1988
From:	Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20150B495	List: IR 05000155/1988009 ML20150B492
References
50-155-88-09, 50-155-88-9, IEB-88-003, IEB-88-3, NUDOCS 8807120048
Download: ML20150B509 (11)

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U.S. NUCLEAR REGULATORY C0FMISSION

REGION III

Report No.~ 50-155/88009(DRP)

Docket No. 50-155 License No. DPR-6 Licensee: Consumers Power Company 212 West Michigan' Avenue Jackson, MI 49201 Facility Name: Big Rock Point Nuclear Plant Inspection At: Charlevoix, MI 49720 Inspection Conducted: April 20 through June 1,1988 Inspectors: S. Guthrie E. Plettner Approved By: . ack kief MN Pr jects Section 28 Date

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Inspection Summary Inspection on April 20 through June 1, 1988 (Report No. 50-155/88009(DRP))

Areas Inspected: Routine, unannounced inspection conducted by the Senior j Resident Inspector. . The functional areas inspected consisted of licensee action on-a previous inspection finding involving site QA staffing, operational safety verification involving day to day plant activities, maintenance observations of work performed on an emergency diesel generator, three digital temperature recorders, a circulating water pump, and several pressure relief valves, security, licensee action in response to NRC Bulletin 88-03, and licensing activities. The inspection invnived a total of 86 inspector-hours by two NRC inspector Results: In general the licensee has demonstrated a desire to respond in a timely manner to issues and concerns presented to them by the NRC. The operational safety maintenance, and security programs appeared adequate to ensure public health and safety. Two open items were identified in the maintenance are Both items involved safety issues with pressure relief valve DR 880629 ADOCK 05000155-PDC

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1.- Persons Contacted

• T. Elward, Plant Superintendent

• G. Petitjean, Planning and Administrative Services Superintendent

• G. Withrow, Engineering Maintenance Superintendent

• R. Alexander. Technical Engineer

, *R.'Abel,' Production and Plant Performance Superintendent

• L. Monshor, Quality Assurance Superintendent R. Barnhart, Senior Quality Assurance Administrator

• P. Donnelly, Senior Review Supervisor, Nuclear Activities Dep D. Swem, Senior Enginear

/ G. Sonnenberg, Materials Services Supervisor D. Staton, Shift Supervisor W. Trubilowicz, Operations Supervisor

• J. Beer, Chemistry / Health Physics Superintendent E. Evans, Senior Engineer R. Brady, Senior Plant Technical Analyst J. Tilton, General Engineer D. Kelly, Maintenance Supervisor D. Ball, Maintenance Supervisor W. Blosh, Maintenance Engineer ,

M. Acker, Senior Engineer J. Toskey, General Engineer L. Darrah, Shif t Supervisor J. Horan, Shift Supervisor R. Scheels, Shift Supervisor J. Warner, Property Protection Supervisor T. Fisher, Senior Quality Assurance Administrator R. Krchmar, General Quality Assurance Analyst R. Boss, Reactor Engineer-

• R. Buckner, Nuclear Training Administrator The inspector 'also contacted other licensee personnel in the Operations, Maintenance, Radiation Protection, and Technical Department * Denotes those present at exit intervie . Licensee Action on Previous Inspection Findings (CLOSED) Open Item 87006-01, Site QA Staffing concerns. The inspector has monitored over a two year period the adequacy of QA/QC staff size and determined that the existing staff size appears adequate to meet the goals of the QA organization. Any future reduction in staff without a corresponding decrease in site staff work load would apparently detract from the site staff's ability to meet the organizational goals and commitment _ - _ _ _ _ _ - - _ _ _-_ ._ .__ _ __ . _ _ _ - - _ _- _

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' Operational Safety Verification (71707)

The inspector observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the containment sphere and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to-verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. During the inspection period, the inspector walked down the accessible portions of the Liquid Poison, Emergency . Condenser, Reactor Depressurization, Post Incident, Core Spray and Containment Spray systems to verify operabilit The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barrelin >

fio violations or deviations were identified in this are . Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specification 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 calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc On April 20 the inspector observed replacement of injectors on the emergency diesel generator by licensee and vendor personne Procedural guidance was utilize On April 20-21 the inspector reviewed the licensee's actions to resolve problems with electrical noise on three digital temperature recorders scheduled to replace original plant equipment. The recorders were scheduled for installation to record plant temperature, reactor vessel and ste6m drum metal temperatures, and

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turbine generator lubricating oil temperatures. Approximately two weeks before scheduled installation, while energized for bench testing in the I l C lab the licensee observed that electrical _ noise generated from the recorders was sufficient to cause significant electrical noise in hand held radios carried through adjacent

' hallways several feet away. The effect of electrical noise generated by these recorders after installation in the relatively small control-room can not be predicted by the licensee or the vendor. The licensee's goal of zero electrical noise caused the cancellation of the modification. Operator and the inspector's observation of the original plant equipment scheduled for replacement indicated that the accuracy of the temperatures indicated was never in question but that wear on moving parts in the recorder's mechanical linkages required regular maintenance to ensure continued multi-point recordin On April 21 the inspector observed the removal of the No. 2 circulating water pump for inspection and repair as require Initial inspection revealed significant erosion of the impeller on this single stage, high volume pump. The worst erosion was observed at the outer edge of one of the two impeller vanes where the impeller had worn to only several thousanths of an inch from its original thickness of approximately one-quarter inch. The pump was inspected after 25 years of service because of the manufacturer's recommended inspection schedul The pump had been inspected approximately 5 years ago with no significant erosion observed. The licensee's vibration analysis program had detected some wear on the pum On May 25 the inspector observed the removal and replacement of pressure relief valve RV-5077. Yhe repair was accomplished under maintenance work request 88-P15-0032 During this inspection period the inspector reviewed the activities of the licensee to address problems with Reactor Depressurization System (RDS) depressurization valves. The depressurization valve is designed to open upon energization of the pilot valve, either manually or automatically in response to specified plant condition Energization of the solenoid valve opens the pilot valve, venting the area above the main piston and permitting system pressure (normally

. 1330 psig) to fully lift the main disc off its sea On April 21 depressurization valve "B" (SV-4985) was tested at a laboratory with unsatisfactory results. The valve was initially verified to be leak free in the test stand when pressurized to 1350 psig. After reducing pressure to 1000 psig for full stroke functional testing, repeated attempts to stroke the valve by energizing the solenoid were unsuccessful and resulted in no valve stem movement although the pilot valve was verified to be actuating properly. In the initial unsuccessful tests the solenoid was energized for approximately two seconds. Licensee data for testing of other valves in 1985 showed the delay time from solenoid energization to comencement of st. , travel to be 600 milliseconds

! for initial lifts and approxinat iy 1500 milliseconds for subsequent l lifts when a two phase steam /wa* ar mixture was present in the discharge piping.

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The licensee then reduced pressure to zero, energized the solenoid and steadily increased pressure in a minimum pressurization tes With zero pressure downstream a steam pressure of 650 psig was required to lift the disc from the sea Inspection revealed the stem and stem guide area to be corroded with a black oxide and the stem displayed drag marks and gouges. The corrosion which had inhibited stem travel during lift tests also prevented reseating of the disc on the seat and testing was terminated. Scheduled replacement of a cracked seat, top assembly rebuild and machining of main disc as. required was connence Initial tests were conducted in the horizontal position although the valve's system orientation is vertical. This testing variable was removed when the laboratory modified the test rig for vertical testin The inspector attended a meeting of the Corrective Action Review Board convened to address the issue. The licensee decided to send a second valve to the testing laboratory for as found testing and to investigate means to accurately assess the condition of the remaining two valve The inspector expressed the following concerns to the licensee on April 2 (1) Based on the unsuccessful as-found testing of SV-4985, one of four RDS trains was apparently inoperable, or at least seriously degraded, for some indeterminable period prior to shutdown on April 9, 1988. The inability of SV-4985 to perform its safety function would seriously degrade the licensee's ability to respond to a plant emergency. Multiple valve failure during a LOCA of the size against which RDS was designed to protect would, according to the licensee's analysis, result in core damag (2) The quarterly surveillance required by Technical Specification 4.1.5 to verify valve operability has apparently been an inadequate substitute for full stroke testing. The surveillance, which discharges a small volume of gas upon actuation, verifies operability by verification of the pressure .

decay of the test gas and by brief loss of full closed indication on the valve's main stem and disc. Full open indication is not received during the test, indicating the stem travels some distance less than full stroke. Stem travel of only 1/8 to 1/4 inch is required to lose closed indication, according to the valve's vendo (3) Based on the findings on SV-4985, the condition of depressurization valves in the A,C, and D trains are in questio .

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(4) Determination of the source of corrosion and the rate of corrosive oxide buildup on the stem surfaces should be evaluated against the apparent inadequacy of the surveillance. This evaluation should be factored into the licensee's pending change to Technical Specifications that would reduce testing activities to full . stroke testing of only one of the four depressurization valves each refueling outag (5) Maintenance practices should be enhanced to ensure crafts person's awareness of the need to verify that the stem travel is not degraded by corrosion. Based on the inspector's observations of previous valve disassembly and inspection activities and interviews with maintenance department supervisors, it is apparent that maintenance personnel do not include, as part of their evaluation'a check for corrosive binding of the stem. The inspector's review of Maintenance Procedure MRDS-2, "Replace, Repair and/or Inspect All or Part of the RDS Depressurizing Valves" performed in September and November 1987, determined that the procedural steps which might have identified stem binding were marked "N/A" by the repairman and not performe (6) The inspector clearly conveyed the expectations of Regional and NRR Management that an adequate licensee response to the inspector's concerns would require testing of all four depressurization valves in a manner that ensures acceptable valve performance and full valve stem travel. Testing approaches other than live stem testing at a laboratory could be acceptable provided the test methodology yielded comparable result On April 29 the inspector attended a meeting of the Corrective Action Review Board convened to examine the failure of SV-4985 and devise a course of action for future testing. The licensee elected to send SV-4984, the "A" train depressurization valve, to the steam laboratory for testin On May 6-9 the inspector reviewed testing of RDS Depressurization Valve SV-4984 performed May 4. The inhibited stem travel observed during SV-4984 testing was not in evidence and the valve operated with a delay time of 260 milliseconds on the initial lift. A second test resulted in a delay time of 880 millisecond However, on the third attempt the valve failed to operate after 35 seconds of solenoid actuation. The licensee was unable to establish an apparent cause. Overheating or other problems with the pilot valve solenoid were apparently ruled out when the licensee immediately perfonned three successful minimum pressurization tests in which only 12 psig was required to lift the main disc off its seat, indicating pilot valve operabilit Repeated testing eventually resulted in pilot valve disc / seat degradation j

and the vendor commenced repairs to SV-493 l-

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In an effort to address the inspector's concerns over the ability of the quarterly surveillance to verify full stroke stem travel as required by applicable codes the licensee designed and constructed a test rig installed on a spare top assembly which employed multiple reed switches operated by movement of the magnetized rod atop the main stem. With the test rig installed on the recently rebuilt and reinstalled SV-4985, a quarterly surveillance as described earlier was satisfactorily perfonned on May 8. Although control room closed position indication was lost and regained, thus satisfying the surveillance's acceptance criteria, strip charts attached to the various test reed switches did not indicate upward stem travel. These results lent support to the inspector's position that stem travel is very slight, possibly in the range of 1/8 to 1/4 inch, and is sufficient only to depressurize the limited volume of gas used as the test medium. Previous performances of the quarterly surveillance indicates that total stem travel time to depressurize the gas is 0.8 - seconds. Based on the inability of the special test to verify full stroke travel the licensee elected to remove depressurization valves SV-4986 and SV-4987 on the "C" and "0" trains for full stroke testing at the laboratory facilit During the period April 28 through May 9 the inspector researched licensing correspondence to assess the licensee's position on full stroke testing of depressurization valves. In the licensee's earliest responses to NRC inquiries regarding the feasibility of full stroke testing the licensee repeatedly stated that the valves could not be full stroke tested without application of continuous primary system pressure, an unacceptable approach because the depressurization valves discharge into the containment environmen In the installed configuration only partial stroke testing is possible. Additional NRC request for the specific technical basis used to determine the inability to full stroke test were met with a restatement of the limitations that prevented full stroke testing in the installed configuration and did not address why partial stroke testing provided by the gas testing described earlier was an acceptable substitute. In the licensee's June 1, 1983 response to Issue No. 83 of the Integrated Assessment (NUREG-0828), the licensee acknowledged that, in their opinion, partial stroke testing "may not adequately verify operability of the system". Noting the licensee staff's position that the present test "only provides a means of verifying a small fraction of total stroke", the licensee's Technical Review Group (TRG)

concluded the issue carried "both safe shutdown and radionuclide release implications". The issue was ranked and included in the Big Rock Point Integrated Plan, also known as the Living Schedul By letter dated March 2, 1984, NRC responded to the licensee's request for specific relief from full stroke testing of depressurization valve Acknowledging the licensee's conclusion that thn partial stroke alternative testing was the only realistic approach to valve testing while

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installed in the system, the staff stated that "it is unacceptable to never full-stroke exercise the valves and that the licensee should investigate alternative methods of testing...". The licensee responded by

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letter October 4,1984, stating only that the issue will continue to be evaluated within the context of the Big Rock Point Integrated Plan. The NRC staff repeated its position in correspondence dated October 15, 1986, which tr @ mitted the Safety Evaluation Report, Pump and Valve Inservice Testing Program. The report identified the applicable Section XI code requirements and repeated the licensee's relief request and the basis for that request, and acknowledged that testing in the installed configuration was realistically limited to the partial stroke gas testing described earlier. The conclusion stated that "never full stroke exercising these valves is unacceptable and the licensee is required to full stroke exercise these valves on a refueling outage frequency". The licensee responded on December 22, 1986 by once again repeating that in the installed configuration full stroke testing was not feasible and that the issue was being investigated as part of the Integrated Pla The inspector presented the following concerns:

(1) The need to full stroke test depressurization valves in order to fully comply with the ASME code for inservice testing was known to the licensee in 1978, based on the December 22, 1978, relief reques (2) The substantial safety significance of the RDS depressurization valves and the licensee staff's concerns over adequacy of the surveillance to verify valve operability has been well documented since June, 198 (3) The NRC staff clearly stated expectation that a means should be developed to conduct full stroke testing have been known to the licensee since March,198 (4) The NRC staff's conclusion that never full stroke testing the valves is unacceptable and the clearly stated requirement that the valves are to be full stroke exercised on a refueling outage frequency have been known to the licensee since October 198 (5) Although the need to full stroke test has been carried on the Integrated Plan for several years, a review of licensee activity indicated virtually no licensee effort toward that en One depressurization valve was tested in 1985 following maintenance performed by the vendor at the test lab, but that testing was to verify post maintenance operability and not part of any program to investigate full stoke test methods or satisfy regulatory or code requirements. A licensee submittal dated March 22, 1988 proposed a change to Technical Specifications that would submit one valve per refueling outage to full stroke testing. The test of SV-4985 was the first in that series of tests although the primary reason for sending the valve to the test laboratory was for seat repairs performed by the vendor.

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On May 31-June 1 the inspector and Regional staff conducted several conference calls with the Plant Superintendent and his staff to discuss the requirements described above and the lack of activity to facilitate full stroke testing. The licensee committed to:

(1) Submit a letter before restart to Region III describing the corrective actions taken to return the valves to service including full stroke testin (2) Submit a Technical Specification change to NRR to full stroke test each valve every refueling ~ outage and to inspect the internal components of one valve every refueling outag These items will be traced as open item 155/88009-0 On May 8 the inspector reviewed testing conducted May 6 on steam drum safety relief valve RV-500 The relief valve was tested in accordance with ANSI /ASME OM-1,1981, "Requirements for Inservice Performance Testina of Nuclear Power Plant Pressure Relief Devices" section INV-3513, which specifies testing of adaitional valves when any valve in a system fails to function properly during regular testing. As described in Report No. 155/88005(DRP), section 3.a.,

the first steam drum relief valve (RV-5000) tested April 14 lifted approximately 5 pounds below the lower setpoint tolerance valu Testing of RV-5001 yielded similar results. Compared to a setpoint of 1545 plus or minus 15 psig, RV-5001 lifted at 1525 on the initial, ,

as-found test. Subsequent lifts were consistent at 1526 and 1527 psig. Like the test on RV-5000 there was no evidence of disc to seat adhesion observed during relief valve testing during previous outage Discussions were conducted between the licensee and Region III staff en functional perfonnance and meeting tolerance values. The licensee committed to Region III to either test the remaining valves or provide an acceptable analysis of why it is not a problem. The licensee requested input from the State of Michigan Boiler Division i Inspector and Authorized Nuclear Inspector to help resolve the problem. The licensee had not received the information at the close of this reporting period. This will be tracked as open item 155/88009-02.

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No violations.or deviations were identified in this this are . Security (71707)

On April 20 the inspector reviewed licensee plans to provide security during removal of a pump from a building within the protected area. The i details of the licensee's compensatory measures were considered safeguards l

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information but appeared appropriate to ensure the integrity of the security barrie On May 9 the inspector reviewed licensee activity to address a deficiency in perimeter electronic detection equipment. The licensee failed to meet the requirements for minimum security staff during a '

period of approximately five hours while compensatory measures were in effect. The deficiency was attributed to supervisory oversigh Details of the event, which were considered Safeguards information, were discussed with regional security specialists. The licensee made the appropriate telephone notifications within one hour after verification that a reportable incident had occurre No violations or deviations were identified in this are . Licensee Action On IE Bulletin 88-03 By it fter dated April 19 the licensca responded within the required 120 days u NRC Bulletin 88-03, Inadequate Latch Engagemert in HFA Type Latching Relays manufactured by General Electric Co. The licensee's response stated that the Big Rock Plant does not use any of the subject relay No violations or deviations were identified in this are . Licensing Activities By letter dated April 19 the licensee informed the staff of corporate personnel and organizational change Those changes include:

Dr. Frederick W. Buckman, fonnerly Senior Vice President, Energy Supply was promoted to President and Chief Ooerating Officer. Mr. Buckman's duties had also included those nonnally carried out by the Vice President, Nuclear Operation Mr. Gordon Heins replaced Dr. Buckman as Senior Vice Presidcat, Fuet 3y Suppl Mr. David Hoffman, fonnerly Palisades Plant General Manager, was named, Vice President, Nuclear Operations. Mr. Hoffman was replaced at Palisades by Mr. Gers ed Slade, formerly Executive Director of '

Nuclear Assuranc During this and the preceding inspection period the inspector verified that the licensee complied with the requirements of '

10CFR19.11.a.4 by posting in a conspicuous location for the required time the Connission's Notice of Violction and Imposition of Civil Pendity. The Violation resulted from a vital area barrier opening is discussed in Inspection Report No. 50-155/87025 (DRSS).

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No violation or deviations-were identified in this are . Open Items Open items are matters which' have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action

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on the part of the NRC or Licensee or both. Open items disclosed during-the inspection are discussed'in Paragraph.4.e and .- Exit Interview-The inspector met with licensee representatives (denoted in Paragraph 1)

throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities. The licensee acknowledged these findings. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes' reviewed by the inspector during the inspectio The licensee did not identify any such documents or processes as proprietary.

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