IR 05000285/1987015
| ML20236G922 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/30/1987 |
| From: | Harrell P, Hunter D, Mullikin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236G911 | List: |
| References | |
| 50-285-87-15, NUDOCS 8708040422 | |
| Download: ML20236G922 (28) | |
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APPENDIX C U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-285/87-15 License: DPR-40 Docket: 50-285 Licensee: Omaha Public Power District (OPPD)
1623 Harney Street Omaha, Nebraska 68102 Facility Name: Fort Calhoun Station Inspection At: Fort Calhoun Station, Blair, Nebraska Inspection Conducted: June 1 through July 15, 1987 Inspector: .P -
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P. H. Harrell, Senior Resident Reactor Dhte Inspector A/ t v/ A R. P. Mul ikin, Project Inspector hk Date '
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Approved: 7 R. Hunter, Chief, Reactor Project Date I ;
5 action B, Reactor Projects Branch 1 Inspection Summan I,r3sjection Cor, ducted June 1 through July 15, 1987 (Report 50-285/87-15)
Areas Inspected: Routine, unannounced inspection including followup on previously identified items, followup on licensee event reports, operational i safety verification, plant tours, safety-related system walkdown, maintenance, f surveillance, security observations, radiological protection observations, i in-office review of periodic and special reports, containment local leak rate- I testing, fire protection / prevention program, observation of activities during plant startup, and followup on items completed prior to plant startu Results: Within the 14 areas insoected, 1 violation (modification of a safety-related system without an approved installation procedure, paragraph 5)
and 1 deviation (failure to revire Procedure ST-ESF-2 as stated in LER 87-001, paragraph 3) were identifi.e h G
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DETAILS
, Persons Contacted
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- Brunnert, Supervisor, Operations Quality Assurance !
- M. Core, Supervisor, Maintenance (Acting Plant Manager)
T. Dexter, Supervisor, Security J. Foley, Supervisor, I&C and Electrical Field Maintenance ;
- J. Gasper, Manager, Administrative and Training Services !
J. Kecy, Acting Reactor Engineer
- L. Kusek, Supervisor, Operations ,
- D. Munderloh, Plant Licensing Engineer l
- T. McIvor, Supervisor, Technical j
- K. Morris, Division Manager, Quality Assurance and Regulatory Affairs R. Mueller, Plant Engineer ]
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- A. Richard, Manager, Quality Assurance ;
- G. Roach, Supervisor, Chemical and Radiation Protection
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S. Willrett, Supervisor, Administrative Services and Security
- Denotes attendance at the monthly exit intervie The NRC inspectors also contacted other plant personnel, including l operators, technicians, and administrative personnel.
I Followup on Previously_ Identified Items 1 (Closed) Deviation 285/8702-01: Administrative control of containment isolation valves.
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The licensee completed changes to procedures and/or drawings to l reflect the required locked or seal-wired shut positions for-all l containment isolation valves. The procedural changes made by the l
licensee should ensure that containment isolation is established and maintained during future plant operation. The basis for the procedure changes made by the licensee to establish containment integrity was a program established by the licensee to verify all appropriate valves were identified. The program included a review of piping and instrumentation diagrams, the Updated Safety Analysi Report, and system walkdown The licensee changed the documentation listed below to complete all necessary changes required to comply with the licensee's commitments made in response to this deviatio . Safety Injection and Containment Spray System (Drawing E-23866-210-130, Sheet 1, Revision 41)
. Waste Disposal System Flow Diagram (Drawing M-6, Revision 30)
. Reactor. Startup Locked Valves (Procedure OI-RC-28-CL-D, l
Revision 51)
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l 1 The NRC inspector reviewed the revision of the documents listed above. Based on this review and a review performed during the .
previous inspection period, it appeared that the licensee had taken appropriate actions to establish an acceptable program for ;
administrative control of containment isolation valve # (Closed) Severity Level V Violation 285/8702-02: Failure to provide adequate controls to avoid the misuse of outdated or inappropriate operator aid l Upon notification by the NRC inspector that operator aids were out of date, the licensee performed a review and replaced all outdated f ;
document j i
The licensee revised the program used for the control of operator f aids to ensure that the appropriate documentation was maintained in !
an up-to-date statu The revisions to the program included the following items:
The document control clerk issues a list of all procedure '
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changes made each day. The shift technical advisor (STA)
compares the list to the list of operator aids posted in various plant areas to determine if an aid had been revise ir an operator sid appeared on the revision list, the STA made a copy of the affected dochment and posted the revised cop . A change was made to the STA shift turnover log (Form FC-163) to j require the STA on each afternoon shift to verify, in writing, that operator aids had been checked and replaced with a revised copy, as appropriat . Procedure S0-0-41, " Control of Temporary Labels, Curves, Notes, or Instructions Attached to Plant Components and Controls," was revised to require that the STA enter the latest revision for each document into S0-0-41 prior to performing the cuarterly document revision verificatio The NRC inspector reviewed the actions taken by the licensee and it appeared that the licensee had established a program to avoid the use of outdated documents. In addition, the NRC inspector performed a review of selected aids to verify that the correct revision was posted on plant components and control No problems were note '
c. (Closed) Open Item 285/8703-01: Review of problems with charging pump circuitr !
l During the loss of an inverter in July 1986, the licensee experienced l problems with the charging pump control circuitry in that the !
operating pump stopped and neither standby pump could be starte 'l The licensee stated that the circuitry would be tested during the refueling outage to determine the cause of the proble I l
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The licensee performed testing on the charging pump circuitry during <
) the refueling cutage in accordance with Maintenance Order (MO) 062918.,
A detailed troubleshooting and testing procedure, approved by the plant review ceuittee (PRC), was attached to the MO to provide specific step-by-step instructions for checking the charging pump circuitr The licensee performed the procedure and determined that the prob 1cm could not be recreated. A series of tests were performed where each j of the three changing pumps was operated and the other two pumps i placed in the standby mode. In all cases, the. charging pumps l
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6perated as designed. Based on the results of the testing, the licensee stated that the problem previously encountered was a one-time problem and that the charging pump circuitry was !
satisf actory for continued plant operatio ]
The NRC inspector reviewed the test procedure and the test results obtained by the licensee. It appeared that the licensee had j adequately tested the circuit,ry and that the results obtained !
adequately supported the licensee's conclusion j i (0 pen) Severity Level IV Viciation 285/8710-07: Failure to establish i a procedure for performance of local leak rate test The licensee revised Pracedure ST-CONT-3, " Containment Isolation Valve Leak Rate Test," to provide new instruction for the testirq of containment isolation valves. The instructions provided by the licensee provided clear, concise cErections fr the technicia performing the tests. The instructions should ensure that the data obtained during the testing will accurately reflect the leak rate status of the valve After issuing a revision to Procedure ST-CONT-3, the 1icensee performad retests on all the affected valves associated with Penetrations M-20, M-22, M-43, and M-44. The licensee-did not perform retests on the valves associated with Penetrations M-8 and M-48. The testing personnel established, through discussions with operations personnel, that the testing valve lineup for Penetrations M-8 and M-48 was the same for the original test as was specified in the revised testing procedure. The retest of all valves i for the penetrations indicated the valves did not exceed their specified leak rate limit The NRC inspector reviewed the changes made by the licensee to Procedure ST-CONT- It appeared that the testing procedure provided by the licensee established proper instructions for testing of the valves. The NRC inspector also performed a review to verify that tid valve lineup for Penetrations M-8 and M-48 was the same during the original test as was specified in the revised testing instruction The NRC inspector could not definitely establish that the valve lineup was not the same; however, the inspector noted that the
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affected valves in the waste system are normally open during plant shutdown conditions. Based on the review performed of the licensees'
actions, it appeared that the licensee had established an adequate program for testing of containment isolation valve This item remains open pending the establishment and implementation l of a training program for performance of leak rate testing as specified in the licensee's response to this violatio . Licensee Event Report Followup Through direct observation, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that deportability requirements were fulfilled, immediato corrective action was accomplished, and corrective action to prevent recurrei ce had been accomplished in accordance with Technical Specifications (TS).
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The following LERs listed below were reviewed, but were not closed pendino completion of licensee actions as described in the discussions provided below:
87-001 - Surveillance test performance error 87-010 - Failure to perform testing on auxiliary feddwater system valves The LERs listed below are closed:
87-004 - Failure to inspect a fire barrier penetration seal 87-007 - Incorrect wiring of a containment post-accident water level transmitter 87-016 - Failure to perform a surveillance test on the Halon system within the required frequency LER 87-001 reported an event where a limiting condition for operat$on (LCO) was entered during the performance of Surveillance Test ST-ESF-2. The test was performed to verify the operability of i an engineered safeguards (ESF) train. The testing instructions i required that a check be performed to verify no ESF equipment in the train not being testing was inoperable prior to removing the other train for testin Contrary to the proceoural requirements, during performance of ST-ESF-2 on January 8,1987, the operator performing the test removed one train of ESF eovipment from service when ESF equipment in the other train was inoperable. This action resulted in an entry into a TS LC To prevent recurrence of this event, the licensee stated in LCR 87-001 that the responsibility for verifying there was no i inoperable ESF equipment in the train not being tested was assigned to the shif t superviso The licensee stated that a revision had l
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Deen made to Drocedure ST-ESF-2 to reflect the change in the responsibility for performance of the tes LER 87-001 was issued by the licensee on February 7, 198 During a followup inspection performed on this event, the NRC inspector noted that testing of the ESF equipment in accordance with !
ST-ESF-2 had been completed on February 12, 198 The procedure used during the performance of this test did not state that the shift supervisor was responsible for verifying no ESF equipment in tne train not being tested wat inoperable. This verification was parformed by an operator. In LER 87-001, the licensee. stated that the assignment of the responsibility had been changed; however, no procedure change had been mad The failure to make a change to ST-ESF-2 prior to performance of the test in February 1987 is an apparent deviation frofd a commitment made to the NRC. (285/8715-01) j The NRC inspector performed an additional review and noted that Procedure ST-ESF-2 was not changed to reflect the change in responsibility for performance of the verification until March 20, 198 Procedure ST-ESF-2 was scheduled to be performed on March 12, ;
1987, but was not performed because the p'lant was in a refueling shutdown. The 15 do not require performance of the test during
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refueling shutdown ]
l The NRC inspector reviewed the station logs to establish whether or i not any ESF equipment in the train not being tested was inoperable ]
during the testing performed on February 12, 1987. It was '
established that no ESF equipment was inoperabl ,
l During review of the event, the NRC inspector noted that the licensee i had not performed a generic review to determine if other surveillance l tests should be changed to assign the responsibility for verification l of equipment status to the shift superviso LER 87-O'01 remains open l pending the completion of a review by the license I
b. LER 87-004 reported an event related to the failure to inspect a temporary fire-barrier seal in accordance with TS requirement The seal wasn't inspected because it had not been identified on the i surveillance test used for inspection of temporary fire-barrier !
seal i The licensee added Penetration Seal 81-E-18 to the surveillance test and inspected the seal. The seal Was found to be fully functiona i Subsequent to the identification of a seal not identified on the surveillance test, the licensee performed Suueillance Test ST-FP-9-F.1, " Fire Barrier Penetrations," in April 198 No other seals were identifie The NRC inspector reviewed the actions taken by the licensee to verify that the actions adequately addressed this event. In ,
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addition, the NRC inspector performed a check of selected temporary seals to verify the seals were included on the surveillance tes No problems were note c. LER 87-007 reported an event where the containment post-accident water level transmitter had been wired incorrectly following repair activities during the 1984 refueling outage. The portion of the transmitter affected was the portion that measured from 25 through 50 percent of the full-scale rang Upon identification of this event-by the licensee, the wiring for the affected range was corrected and a test performed to verify that the level transmitter operated correctly. The test was performed using revised Procedure CP-388, " Containment Postaccident Water Level -
Channel 388." The revision to Procedure CP-388 included specific requirements that the technician verify the operability of each range of the level transmitte The NRC inspector reviewed the licensee's actions described above in response to this event. The NRC inspector also reviewed :
Procedure CP-387, " Containment Postaccident Water Level -
Channel 387," to verify that the specific instructions provided in ;
Procedure CP-388 had also been included in Procedure CP-38 No ,
problems were note d. LER 87-010 reported an event related to the failure of the. licensee to include check valves in the auxiliary feedwater (AFW) system in the inservice inspection (ISI) program. The valves (FW-173 and FW-174) were located in the discharge lines of the motor-driven (FW-6) and steam-driven (FW-10) AFW pumps, respectivel As a short-term measure, the licensee took action to verify that the 4 check valves were functional. For verification of the operability of l Valve FW-173, Pump FW-6 was operated during plant .startup to supply )
feedwater to the steam generators. Operation of Pump FW-6 verified the operability of~ Valve FW-173 in accordance with the requirements stated in Procedure ST-FW-1-F.2, " Pump and Remotely Operated Valve Check." For verification of the operability of Valve FW-174, the valve was disassembled in accordance with Procedure ST-ISI-AFW-1-F.1,
"FW-174 Operability Verification," and verified to function normall This verification was performed prior to startup from the, refueling outag The NRC inspector reviewed the results of the operability checks performed for Valves FW-173 and FW-174 in accordance with Procedures ST-FW-1-F.2 and ST-ISI-AFW-1-F.1. Based on this review, it appeared that the licensee had adequately verified the operability of FW-173 and FW-174.
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l In addition to the above short-term measures, the liceasee stated in !
the LER that long-term measures would also be performed. Th !
long-term measures included the following:
. Submission of a request to, and approval by, the NRC's Office of Nuclear Reactor Regulation to change the ISI program to allow testing of the valves during each refueling outage instead of j the currently required frequency of quarterl !
. Perform an independent review of the ISI program to verify that .
all plant components are tested or that an NRC-approved exemption has been issued for those components not tested in accordance with che ISI requirement j
. Perform a review of all procedures related to ISI activities to verify the procedures specifically define what components are - ;
being tested and how the test verifies operability, i LER 87-010 remains open oending the completion of all the long-term actions identified by the licensee in response to this even e. LER 87-016 was issued to report an event where Surveillance Test ST-FP-10 was not conducted in accordance with TS requirement ,
Procedure ST-FP-10 reouired a check be performed to verify the l pressure and weight of the Halon fire protection system storage '
tank A review performed by the licensee indicated confusion 'sted among the two groups designated the responsibility for perfo ace of the test as to which group was responsible for performing ST-FP-10. Due to the confusion, one group thought the other group had performed the surveillance. To eliminate any future confusion as to which group was assigned the prime responsibility for performance of the test, the licensee changed ST-FP-10 to designate the responsibility for '
performance to a single grou The NRC inspector reviewed the change made to ST-FP-10 and verified I that the licensee had designated one' specific group as having primary l responsibility for test performanc The NRC inspector also reviewed j Procedure 50-G-23, " Surveillance Test Program," to determine what 1 instructions existed in the procedure regarding assignment of responsibilities for surveillance test performanc Procedure 50-G-23 stated that the group designated first was ( responsible for test performance when multidisciplinary groups were identified in a surveillance procedure. It appeared that Procedure 50-G-23 provided appropriate instructions for controlling the performance of surveillance test The licensee issued a memo to the appropriate personnel to stress that Procedure 50-G-23 provided instructions for designation of the group responsible for performance of surveillance tests.
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4. 0,perational Safety Verification The NRC inspector cor. ducted reviews and observations of selected activities to verify that facility operations were performed in conformance with the requirements established under 10 CFR, administrative procedures, and the TS. The NRC inspector made several control room observations to verify the following:
. Proper shift staffing
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. Operator adherence to approved procedures and TS requirements )
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. Logs, records, recorder traces, annunciators, panel indications, and j switch positions complied with the appropriate requirements 'l
. Proper return to service of components
. M0s initiated for equipment in need of maintenance ~
. Appropriate conduct of control room and other licensed operators
. Management personnel toured the control room on a regular basis No violations or deviations were identified, I
5. Plant Tours The NRC inspector conducted plant tours at various times to assess plant and equipment conditions. The following items were observed during the ;
tours:
. General plant conditions, including operability of standby equipment, were satisfactory.
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l . Equipment was being maintained in proper condition, without fluid l leaks and excessive vibratio Plant housekeeping and cleanliness practices were observed, including no fire hazards and the control of combustible materia . Performance of work activities was in accordance with approved procedure . Portable gas cylinders were properly stored to prevent pcssible missile hazards.
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. Management personnel toured the operating spaces on a regular basi During plant tours, the NRC inspector noted the following items: 1
. Evidence of smoking was found in the emergency diesel generator rooms These safety-related equipment rooms were designated as NO ,
SM0 KING area !
. Gas cylinders were found not to be secured in storage racks on five separate occasions during this inspection period. One of the unsecured cylinders was in service at the time of discovery. In addition, two gas cylinders were found that had been secured to plant piping and electrical conduits. In the two cases identified, the cylinders were not secured to safety-related piping or conduits. The NRC inspector stressed to licensee personnel, the need to ensure that I cylinders are not secured to safety-related piping or conduit !
. Handrails were removed from the platform above the main steam and feedwater piping for access to the piping during the refueling outag Maintenance personnel did not replace the handrails after the outage was complete The absence of the handrails posed a falling hazard to personnel working in the are On June 3, 1987, the NRC inspector notified the licensee of the missing handrail The licensee replaced the handrails on June 25, 198 . On five occasions during this inspection period, it was noted that j three different fire doors in the plant were not latche The i unlatched fire doors caused the associated fire barriers to be nonfunctional. No determination could be made as to how long the doors were unlatched. In each case, the NRC inspector secured the fire door . Polyethylene bags of tools and waste from the refueling outage had accumulated in Room 59 and in the area adjacent to the personnel air lock. These areas required additional housekeeping attention by the licensee. By the end of this inspection period, the licensee had not provided the additional housekeeping attentio During a tour of the piping penetration area on June 16, 1987, the NRC inspector noted a pressure gage and tubing installed in the. containment sump discharge line, a Class 2 piping system, that was not shown on any plant drawing. At the time of discovery, the plant was at 100 percent power. The gage was installed between the containment isolation valves (HCV-506A and HCV-5068) that were located outside reactor containmen The NRC inspector noted that, under an identified scenario, the pressure gage and tubing could become a part of the containment pressure barrie If, during an event that required containment isolation, the inside valve (HCV-506A) failed to shut and the the outside valve (HCV-506B) did shut, containment isolation would be established. In this case, the gage and tubing would be subjected to containment pressure. Upon notification by the NRC inspector, the licensee removed the gage.
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The NRC inspector requested the licensee provide the documentation that provided instructions for installation of the gagt. After reviewing various types of documentation (e.g , M0s, surveillance tests, calibration procedures, etc.), the licensee could not identify any approved documentation that provided instructions for installation of the tubing and gage. In addition, the licensee could not establish that the tubing and gage were materials approved for installation in a Class 2 piping system by the licensee's quality assurance progra Criterion V of Appendix B to 10 CFR Part 50 states that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawing Section 5.1, " Control of Plant Design and Modifications," of the licensee's Quality Assurance Manual and Procedure S0-G-21, " Station Modification Control," have been established to implement the requirements of Appendix B and require that modification of safety-related systems be
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performed in accordance with written procedures.
l The failure of the licensee to provide appropriate instructions for the modification of the containment sump discharge line by the addition of a pressure gage and tubing is an apparent violation. (285/8517-02)
6. Safety-Related System Walkdown The NRC inspector walked down accessible portions of the following safety-related system to verify systam operability. Operability was determined by verification of selected valve and switch position The system was walked down using the drawings and procedure note Auxiliary feedwater system (Procedure FW-4-CL-A, Revision 30, and Drawings M-253, Revision 56, and M-254, Revision 62)
During the walkdown, the NRC inspector noted no discrepancies between the drawings, procedure, and plant as-built conditions for the selected areas checke No violations or deviations were identifie . Monthly Maintenance Observations The NRC inspector reviewed and/or observed selected station maintenance activities on safety-related systems and components t0 verify the maintenance was conducted in accordance with approved procedures, regulatory requirements, and the TS. The following items were considered during the reviews and/or observations:
The TS limiting conditions for operation were met while systems or components were removed from servic _
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. Approvals were obtained prior to initiating the wor ;
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. Activities were accomplished using approved M0s and were inspected, as applicabl . Functional. testing and/or calibrations were performed prior to returning components or systems to servic . Quality control records were maintaine . Activities were accomplished by qualified personne . Parts and materials used were properly certifie ,
. Radiological and fire prevention controls were implemente The NRC inspector reviewed and/or observed the following maintenance activities:
. Installation of temporary lead shielding (M0 872889)
. Troubleshooting of the charging pump circuitry (M0 862918)
No violations or deviations were identifie . Monthly Surveillance Observations The NRC inspector observed selected portions of the performance of and/or reviewed completed documentation for the TS-required surveillance testing on safety-related systems and components. The NRC inspector verified the following items during the testing:
l . Testing was performed by qualified personnel using approved procedure l l
l . Yest instrumentation was calibrate j
. The TS limiting conditions for operation were me l
. Removal and restoration of the affected system and/or component were !
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. Test results conformed with TS and procedure requirement . Test results were reviewed by personnel other than the individual directing the test.
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. Deficiencies identified during the testing were properly reviewed and *
resolved by appropriate management personnel.
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The NRC inspector observed and/or reviewed the documentation for the following surveillance test activities. The procedures used for the test activities are noted in parenthesi . Monthly emergency diesel generctor operability verification (ST-ESF-6-F.2)
. Containment pressure channel check (ST-ESF-3-F.1)
. Pressurizer pressure channel check (ST-ESF-1-F.2)
. Auxiliary feedwater pump and remotely-operated valve check (ST-FW-1-F.2)
. Inservice inspection of raw water pumps (ST-ISI-PW-3-F.1)
. Automatic load sequencer check (ST-ESF-5-F.1) !
. Calibration of a toxic gas monitor (CP-6287A)
. Flowrate test for diesel-driven fire pump (ST-FP-4)
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No violations or deviations were identifie . Security Observations The NRC inspector verified the physical senrity plan was being implemented by selected observation of the following items:
. The security organization was properly manne . Personnel within the protected area (PA) displayed their .
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identification badge . Vehicles were properly authorized, searched, and escorted or controlled within the P . Persons and packages were properly cleared and checked before entry into the PA was permitte . The effectiveness of the security program was maintained when security equipment failure or impairment required compensatory measures to be employe . The PA barrier was maintained and the ' solation zone kept free of transient materia .
The vital area barriers were maintained and not compromised by breaches or weaknesse :
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. Illumination in the PA was adequate to observe the appropriate areas at nigh . Security monitors at the secondary and central alarm stations were functioning properly for assessment of possible intrusion j i
No violations or deviations were identifie . Radiological Protection Observations The NRC inspector verified that selected activities of the licensee's j radiological protection program were implemented in conformance with the j facility policies and procedures and in compliance with regulatory ;
requirements. The activities listed below were observed and/or reviewed: /
. Health physics (HP) supervisory personnel conducted plant tours to check on activities in progres ]
. Radiation work permits contained the appropriate information to ensure work was performed in a safe and controlled manne . Personnel in radiation controlled areas (RCA) were wearing the j required personnel monitoring equipment and protective clothin . Radiation and/or contaminated areas were properly posted and !
controlled based on the activity levels within the are j
. Personnel properly frisked prior to exiting an RC No violations or deviations were identifie '
11. In-office Review of Periodic and Special Reports In-office review of. periodic and special reports was performed by the NRC resident inspector and/or the Fort Calhoun project inspector to verify the following, as appropriate:
. Reports included the information required by appropriate NRC requirement . Test results and supporting information were consistent with design predictions and specification . Determination that planned corrective actions were adequate for resolution of identified problems.
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. Determination as to whether any information contained in the report should be classified as an abnormal occurrence.
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The NRC inspectors reviewed the following:
. Special reports on inoperability of fire barriers, dated June 5, 15, and 22, 1987
. Clarification of Cycle 11 information, dated June 10, 1987
. Monthly operations report for May 1987, undated
. May 1987 monthly operating report, dated June 12, 1987
. Request for exemption and revision to the inservice inspection plan for testing of the auxiliary feedwater system pump check valves, dated July 1, 1987 During review of reports, NRC personnel identified 10 CFR Part 21 reports submitted by suppliers or vendors that appeared to be applicable to the licensee's facility. The NRC resident inspector provided copies of these reports to the plant licensing engineer for review of applicability by the licensee. The reports provided are listed below:
. A letter dated April 27, 1987, from the SOR Company related to gas bubble formation in pressure switches
. A litter dated September 5, 1986, from the Bechtel Power Corporation related to pipe support tolerance and installation procedures
. A letter dated April 29, 1987, from the Basler Electric Company related to cracking of 0-rings in emergency diesel generator exciters No violations or deviations were identifie . Containment Local Leak Rate Testing The NRC inspector performed this inspection to verify that the local leak rate test program for testing of containment isolation valves and )
penetrations (mechanical and electrical) was performed in accordance with TS requirement The completed testing procedures reviewed are listed below:
. Electrical Penetration Leak Rate Test (ST-CONT-2-F.5)
. Mechanical Penetration Seals . Leak Rate Test (ST-CONT-2-F.6) l
. ContainmentisolationValveLeakRateTest(ST-CONT-3-F.1)
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Personnel Access Lock (PAL) 0-Ring Seat Test (ST-CONT-2-F.1)
. Containment Purge Isolation Valves Leak Rate Test (ST-CONT-3-F.4)
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. Full Transfer Tube Leak Rate Test (ST-CONT-2-F.4)
The procedures listed above were reviewed for the following elements:
. All valves and penetrations were tested in accordance with TS ,
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. Leaking penetrations and valves were repaire . Retests were performed for the valves and penetrations that required repair after initial tests were performe . The test equipment used was in calibration at the time of the tes During perforttiance of the leak rate tests, the licensee noted various valves and mechanical penetrations that leaked and required repair. The licensee repaired the valves and penetrations and retested the repaired components. The retests indicated that the repairs were adequat Four penetrations that required repair were the mechanical penetration
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assemblies for the two main steam and the two feedwater lines. During testing, the licensee identified leaks in the expansion bellows in the penetration assemblie The licensee replaced the bellows in accordance with Procedure MR-FC-87-1 This procedure provided instructions for grinding out the old bellows and welding in a replacement. The NRC inspector reviewed selected portions of the bellows replacement procedure and noted no problem In addition, the NRC inspector reviewed M0s associated with repairs of q penetrations and containment isolation valves. The M0s are listed below:
. Repair of Valve HCV-2504A (M0 870386) l
. Repair of Valve VD-504 (M0 871075)
i . Repair of Valve HCV-383-3 (M0 872442)
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. Repair of Penetration M-88 (M0 872781)
. Repair of Penetration M-87 (M0 872771)
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The NRC inspector reviewed the actions taken and documentation completed by the licensee as discussed above. No problems were note l l l No violations or deviations were identifie i 13. Fire Protection / Prevention Program I The purpose of this portion of this inspection was to verify.that the licensee had implemented a program for fire protection and prevention that i conformed with regulatory requirements, TS, and industry guides and l standard i
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The NRC inspector reviewed the following administrative procedures and verified the procedures adequately implemented the licensee's fire protection program:
. Housekeeping (50-G-6, Revision 18) l i
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. Storage of Critical Element 'and Radioactive Material Packaging, Fire Protection Material, and Limited CQE (S0-G-22, Revision 24)
. Station Fire Protection Plan (50-G-28, Revision 10) l
. Fire Barrier Protection (50-G-58, Revision 3) i
. Fire Prevention During Flame Cutting and Welding Operations (50-M-9, Rcvision 9)
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Fire Watch Duties and Turnover Procedures (S0-0-38, Revision 1)
The NRC inspector also reviewed surveillance procedures and selected completed procedures on fire suppression, detection, and support equipment. The follwing surveillance procedures were reviewed:
. Station Batteries (ST-DC-1, Revision 33)
. Emergency Lighting (ST-DC-4, Revision 7)
. . Fire Detector System (ST-FD-1, Revision 27)
. Battery-Powered Smoke Detectors (ST-FD-2, Revision 3)
. Fire Protection (ST-FP-1, Revision 5) j
. Fire Protection System Diesel Fire Pump Battery (ST-FP-2, Revision 4)- j
. Test of Fire Protection Water Suppression System (ST-FP-3, Revision.9) l
. Full Flow Test of Fire Pumps (ST-FP-4, Revision 5)
. Fire Protection - Auxiliary Building Sprinkler Systems Testing ;
(ST-FP-5, Revision 3). ;
. Fire Protection - Fire Hose Testing (ST-FP-6, Revision 6) ;
. Diesel Fire Pump Surveillance (ST-FP-7, Revision 6)
. FirePumpStrainerInspection(ST-FP-8, Revision 8)
. Fire Barrier Penetrations (ST-FP-9, Revision 10)
. Halon Systems Surveillance (ST-FP-10, Revision 9) *
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. Fire Protection - Functional Test of Diesel Generator Rooms Dry Pipe Deluge Valve (ST-FP-11, Revision 1)
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The NRC inspector toured accessible areas of the plant to review general area conditions, work activities in progress, and visual condition of fire i protection systems and equipmen Combustible materials, flammable and combustible liquids, and gas usage were restricted or properly controlled !
in areas containing safety-related systems and components. Items checked !
included positions of selected valves, fire barrier conditions, hose stations, hose houses, Halon system lineups, fire lockers, and fire extinguishers for type, location, and condition. During the tour, . it was discovered that the Halon fire suppression system for the cable spreading room was inoperable. The NRC inspector verified that a continuous firewatch was in effect in this room as required by the TS. Also, it was ;
found that the hot-leg temperature indicator on the alternate shutdown !
panel (AI-185) was inoperabl The licensee was notified and MO 873525 was initiated to repair the temperature indicato An inspection of the upper electrical penetration room revealed that an oily substance, determined by the licensee to be tendon grease, was being secreted through the seismic gap fire seal at the northwest corner of the-room. This seal is located where the' auxiliary building joins the containment building and is a barrier between the radiologically controlled and uncontrolled areas of the auxiliary building. The outer caulking on the seal was destroyed on a portion of the gap, but it could not be determined if this was due to the grease. This item remains unresolved pending the licensee's investigation as to the effect of the tendon grease on the seismic gap fire seals. (285/8715-03) -j There was no welding, cutting, or use of open flame ignition sources found in the areas toured. General housekeeping conditions were found to be in need of additional licensee attention. There were no construction activities in progress in the toured area Fire protection systems and equipment installed for protection of safety-related areas were found to be functional (except for the Halon system mentioned above). Fire brigade equipment, including emergency 1 breathing apparatus, was found to be properly stored and maintaine )
i The NRC inspector reviewed fire brigade training and drill records and I reviewed the current roster of qualified brigade members. The records !
were in order and confirmed that training and drills were being conducted l at the specified interval However, it was stated by the licensee that i the use of firefighting training in confined spaces was not part of their regular training activities. The licensee should consider including this ;
training as a permanent part of their annual fire brigade trainin No violations or deviations were identifie J l
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14. Observation of Activities During Plant Startup The NRC inspector reviewed a variety of activities associated with startup I of the plant from the refueling outage. The reviews were performed to l verify plant startup was conducted in accordance with the TS, operating I procedures, and administrative procedure During the performance of this portion of this inspection, the NRC inspector observed and/or reviewed selected activities in the following areas:
. Determination of the reactor shutdown margin was properly performed and a program established for determination of the shutdown margin during the next cycl . Plant systems were re' turned to service prior to plant startu . Plant startup, heatup, and approach to criticality were conducted in accordance with approved procedure . Operational activitics were evaluated to verify operators were attentive and responsive to plant parameters and conditions, procedures were used and followed, equipment changes were appropriately documented, and plant operating conditions were effectively monitore . Data from core physics tests were reviewed to verify that the data met the established acceptance criteri !
. Calculation of core thermal power was performed using an adequate procedur . The nuclear instrumentation systems were calibrate . The worth of the control rods was properly calculated and the procedure used conformed to licensee commitment . The value determined for the moderator temperature coefficient was consistent with the predicted values and the TS requirement . The total power coefficient of reactivity measurements complied with i TS requirements, were performed in accordance with licensee j commitments, and were conducted in accordance with the established i procedure, The NRC inspectcr reviewed the procedures listed below to verify proper completion of the items listed above:
. Post Refueling Core Physics Testing and Power Ascension (SP-PRCPT-1)
. Core Thermal Power Calculation NSSS Calorimetric (SP-CTPC-1) $
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. Radial Peaking Factors (ST-ICI-1)
. Peactor Coolant System Flowrate (ST-RCF-1)
. Reactor Startup (OP-7)
. CEA Drive System Interlocks Check (ST-CEA-1-F.1)
. Power Range Safety Channels (ST-RPS-1-F.2)
. Incore/Excore Calibration (ST-RPS-13) l
. Fuel Assembly Uplift Condition Detection (SP-FAUD-1)
. Control Element Assemblies Group Indication Light Check and CEA Drop Times (ST-CEA-1-F.9)
. Shutdown Margin Verification (ST-SDM-1-F.1)
Based on the review performed by the NRC inspector of selected activities, it appeared that the licensee performed the plant startup from the refueling outage in accordance with the appropriate requirement No violations or deviations were identified.
15. Followup on Items Completed Prior to Plant Startup Reviews were performed by NRC inspectors in the areas of maintenance and welding in March and April 1987. The details of the inspections are l provided in NRC Inspection Reports 50-285/87-05 and 50-285/87-08. During these inspections, several major problems were identified and a subsequent enforcement conference was held with the licensee in the NRC Region IV offices on May 14, 1987, to discuss the issues related tc the inspection l During this conference, the licensee committed to perform certain l short-term corrective actions prior to plant startup from the refueling outage. In a letter, dated May 22, 1987, from the licensee to the NRC, a list of the commitments was provide l j
The purpose of this portion of this inspection was to verify that the l licensee had completed the short-term corrective actions.as stated in the !
letter. The NRC inspector reviewed selected portions of each corrective action taken by the licensee to verify the action had been completed. A list of the items and the review performed by the NRC inspector is provided below: The System Acceptance Committee (SAC) reviewed and accepted 1 installation of the seismic supports for masonry wells installed in l accordance with Modification Request (MR) FC-81-16 Tne NRC inspector reviewed the results of the SAC acceptance of l MR-FC-81-180 and determined that the SAC approved the installation of the seismic supports on May 26, 1987.
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During the review by the SAC, discrepancies with the MR-FC-81-180 were noted. The discrepancies were documented in Operating Incidents (01) 2203 and 2727. The discrepancies were related to installation of concrete anchors, inadequate welding' instructions, and other installation inadequacies. These discrepancies were resolved prior to startup of the plan Based on the documentation a rev~iewed, it appeared that the SAC had properly reviewed and accepted MR-FC-81-18 Subsequent to SAC acceotance, a review was performed of portions of MR-FC-81-180'and problems were noted with the installation of the seismic wall supports. The details of the review are provided in NRC Intpection Report 50-285/87-13. No items icentified in the report were required to be completed prior to plant startu b. Perform the items listed below for repair of the emergency feedwater storage tank (EFWST).
(1) Welds were properly repaired and weld inspection was performe This item was verified to be completed during an inspection performed by a Region IV inspector. The results of the review is detailed in NRC Inspection Report 50-285/87-1 (2) The EFWST was sandblasted and coate This item was completed in accordance with instructions provided in MO 862360. A detailed procedure, approved by the plant review committee (PRC), was attached to the M0 and provided detailed step-by-step instructions for sandblasting and coating the EFWS Based on review of the completed documentation and field inspections by the NRC inspector, it appeared that the licensee properly sandblasted and coated the EFWS (3) Hydrostatic testing of the EFWST was performe The EFWST was hydrostatically tested in accordance with a PRC-approved procedure attached to M0 872077. The hydrostatic
) test was successfully performed at the correct test pressure ,
' with quality control verification of no leakage during the tes Based on a review of select portions of the completed documentation by the NRC inspector, it appeared that the EFWST was properly hydrostatically teste c. Welding performed on critical quality elements during the 1987 refueling outage was performed by qualified welders using qualified welding procedure A review of welder qualifications for welds installed dur:ng the outage was performed by the licensee. During the review,.the licensee noted that one welder had installed a weld on a seismic pipe
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hanger when the welder was not qualified at the time the weld was mad The licensee issued 0I 2774 to document the problem and the subsequent resolution of the problem. The licensee performed an engineering evaluation of the weld and determined that the weld was technically acceptabl The NRC inspector reviewed OI 2774 and the associated documentatio Based on a review of the documentation, it appeared that the licensee adequately addressed and resolved the identified welding proble The main steam safety valves (MSSV) were adequately installed using new fasteners and gaskets, and the proper torque value The licensee removed and reinstalled all 10 MSSVs to ensure that the MSSVs were properly installed using new gaskets and fasteners, and that the MSSVs were installed to the proper torque. The licensee installed the MSSVs using the instructions provided in Maintenance Procedure MP-MS-4. The method used to apply torque to the fasteners was the bolt stretch method. Procedure MP-MS-4 was reviewed and approved for use by the PRC. For each of the 10 MSSVs, the licensee issued an M0 and attached Procedure MP-MS-4 to the M0 to establish a record of the completion of the maintenance activities, i l
The NRC inspector reviewed a selected number of the completed M0s for review. The M0s reviewed included M0 872743 for MS-279, M0 872501 for MS-275, MO 872502 for MS-276, and MO 872503 for MS-277. During review of the M0s, no problems were note The pressurizer safety valves (PSV) were adequately installed and a review was performed to verify the proper amount of torque was use The licensee performed a formal calculation to verify that the PSVs were properly installed using the correct amount of torque on the valve flange fasteners. The calculation indicated that the valve fasteners should be installed at a torque of 750 foot pounds. The licensee specified the value of 750 foot pounds be used for installation of the PSVs in Maintenance Procedure MP-PSV-1- The NRC inspector reviewed the calculation prepared by the licensee and Procedure MP-PSV-1-3 to verify that they had been properly complete No problems were note Limitorque operators on valves that were included in the electrical equipment qualification program had the mixture of two different greases removed and replaced by an approved lubrican J The licensee issued Maintenance Procedure MP-DEGRE-1 and attached M0s to provide instructions for removal of the grease mixture in the limitorque operators and replacement with an approved lubricant. To remove the grease mixture, the licensee's procecure required that the Limitorque operator be disassembled and the grease removed by
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physically cleaning with a cloth. After removal of the grease
. mixture, the operators were' inspected for worn or broken parts. Any defective parts were replaced. The operator was reassembled and then regreased using Exxon' Nebula EP-0 greas i The NRC inspector reviewed Procedure MP-DEGRE-1 to verify that' "
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adequate instructions had been provided. .The inspector also reviewed selected M0s issued for grease removal and replacement. The ;
completed M0s reviewed included M0 863589 for Valve HCV-314 and - >
M0 863590 for Valve HCV-315. Based on the results of this review, it j
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appeared that the licensee had taken appropriate actions to remove the greu e mixture and regrease the valve operators with an approved i lubrican !
~ The containment vent f ans were properly installed by accomplishment of the items listed below: ,
f (1) Tne insulation resistance of the fan power cables was tested (meggered) to verify the cables were in satisfactory conditio The licensee meggered the containment vent fans in accordance l with the instructions provided in newly issued Special Procedure SP-EE-MEGGER. The licensee had not previously meggered the: vent fan cables. The results of the-megger testing ;
performed by the licensee indicated that the power cables to the
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fans were satisfactory for ' continued operatio The NRC inspector reviewed the results votained by the licensee during performance of Procedure SP-EE ,JGGER to verify that the procedure was properly completed and that'the cables for 'all q fans had been tested. The inspector noted no problems during )
the revie j (2) Verify the value of the torque used for fan blade bolting and the angle of the fan blades were correct'
The licensee revised Maintenance Procedures MP-VA-3 and MP-VA-7 to include requirements for the field checking of the.value of the torque used on the fan blade bolting and measurement of the <
fan blade angle. For each fan checked, Procedure.MP-VA-3 or MP-VA-7, as appropriate, was attached to'an'M0 issued for each specific fan.~ The review performed by'the' licensee indicated !
that the fan bolting'had previously been installed at.the correct torque. Some fan blades on each fan ware readjusted b the licensee to the valve ~specified in revised Procedures MP-VA-3 and MP-VA- ,
The NRC inspector reviewed M0 872746 for Fan VA-3A, MO 872747 for Fan VA-3B, MG 872748 for Fan-VA-7C,'and MO 872749 for Fan VA-7D' to verify that procedures MP-VA-3 and MP-VA-7.had been properly complete No problems'were noted during the revie _ _ _ _ _
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24 M0s written during the 1987 refueling outage were reviewed to verify that torque values had been provided, if appropriate, for maintenance activities performed on safety-related system The licensee performed a review of all M0s and determined that a total of 181 M0s completed during the outage required torque values be provided. A review indicated that, of the total of 181, 41 had the values specified. Of the rema.ning 140 M0s, the value of the torque was checked and verified to t,e correct for 29 M0s. ~9r the remaining 111 M0s, the licensee performed a justification ft continued operation (JC0) to ascertain that the equipment ar.. dor components would perform their intended safety function The licensee is in the process of establishing a permanent program for determining the appropriate value of torque to be used fcr each plant application. In the interim, the licensee implemented a temporary program where all MOs are routed through the plant engineering section for a determination of a torque value, if appropriate. The temporary program was established based on guidance recently published in the EPRI book, " Good Bolting Practices Manual."
The NRC inspector reviewed selected M0s to verify that the licensee had adequately addressed the need for providing torque values. The inspector reviewed the JCOs prepared by the licensee for a variety of maintenance activities. JC0s were reviewed for M0 857717 (installation of a new synchronization switch for an emergency diesel generator), MO 871159 (tightening of wire connecting screws in an electrical junction box), M0 870953 (replacement of the detector in a radiation monitor), MD 845340 (replacement of the sleeve bushings in a 4160-volt breaker), and M0 871929 (maintenance on the nonpressure retaining boundaries of Limitorque valve operators). The NRC inspector also reviewed selected design change installation packages in the mechanical, electrical, and civil disciplines and noted that torque values were provided, when appropriat Based on the review performed by the NRC inspector, it appeared that the licensee had performed an appropriate review of M0s and had adequately addressed the need to supply the value of torques used in safety-related system l
The reviews discussed above were performed to verify that the licensee had adequately implemented actions to address the short-term corrective j actions for the problems identified in NRC Inspection Reports 50-285/87-05 and 50-285/87-08. The long-term corrective actions will be reviewed during followup of the licensee's response to the violations identified in
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No violations or deviations were identifie I l
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16. Unresolved Item j l
An unresolved item is a matter about which more information is required in l order to determine whether it is acceptable, a violation, or a deviatio One unresolved item is discussed is paragraph 1 Item ParagrapF; Subject j 285/8715-03 13 Effect of tendon grease on-seismic gap fire seals 17. Exit Interview The NRC inspectrar met with Mr. M. R. Core (Acting Plant Manager) and other members of the licensee staff at the end of this inspectio At this meeting, the NRC inspector summarized the scope of the inspection and the finding !
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