IR 05000443/1989008
| ML20247K961 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/13/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20247K954 | List: |
| References | |
| 50-443-89-08, 50-443-89-8, IEIN-83-56, IEIN-84-42, NUDOCS 8909220080 | |
| Download: ML20247K961 (18) | |
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II. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket / Report No.: 50-443/89-08 License No.: NPF-67 Licensee:
Public Service Company of.New Hampshire 1000 Elm Street Manchester, New Hampshire 03105'
Facility:
Seabrook Station, Unit No.1, Seabrook, New Hampshire Dates:
July 1 - August 17, 1989 Inspectors:
N. Dudley, Senior Resident Inspector J. Petrosino, Vendor Inspection Branch, NRR l Approved By:
1 O, bM, h 9l13h Ebe C. McCabe, Chief, Reactor Projects Section 3B Date
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Azeas Inspected: Operational safety, maintenance, surveillance, security inci-dent repot ts, reportable events, allegations, and open items.
Results:
The problems associated with operational procedures controlling effluent re-leases'from facility sumps and tanks is a recurring weakness (paragraphs 3.3 and 6.1).
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Maintenance and surveillance activities were found to.be well managed and con-trolled, with the exception of the repair of a snubber on the diesel generator service water pipe (paragraph 6.3).
l The licensee demonstrated a commitment to properly disposition a questionable I
chemistry surveillance (paragraph 6.2) and in the tracking of documentation for offsite shipment of potentially contaminated components (paragraph 5.2).
Quality control provided effective follow-up on problems involving the use of non qualified duct tape in the primary auxiliary building (paragraph 4.2).
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8909220080 890913 PDR ADOCK 05000443 Q
FDC
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T TABLE OF CONTENTS PAGE l
1.
Persons Contacted.................................
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L 2.
Summa ry of NRC and Faci l i ty Acti vi ti e s...............................
3.
Status of Previous Inspection Findings (71707, 92701, 92702).........
2-3.1 (Closed) UNR 87-16-02; RHR Weld Failure........................
3.2 (Closed) NOV 89-03-02; Boron Dilution Flowpath Valves Found Mispositioned.................................................
3.3 (Closed) UNR 89-06-01; Release of Liquid Waste Tank.............
5-3.4 (Closed) UNR 89-80-01;.Use of Step 11sts to Bypass Normal Review Cyc1e..........................................................
3.5 Painting Schedule for PAB F1oors................................
3.6 Fo u r Ho ur Eme rge n cy Li g ht i n g....................................
3.7 Nuclear Safety Audit and Review Committee......................
-6 3.8 Service Water Supply Valves SW-18 Failure.......................
3.9 Main Steam Isolation Valve Actuators............................
4.
Ope rati o nal - Sa f ety ( 71707 )..........................................
4.1 Plant Operations Review.........................................
4.2 Use of Unqualified Tape in PAB..................................
8-5.
Security (71707)....................................................
8-5.1 Review of Internal Security Incident Reports....................
5.2 Transfer of Special Nuclear Materia 1............................
9-6.
Plant Operations (71707,93702)......................................
6.1 Unmonitored Release from the Turbine Building Sump..............
6.2 Investigation of Chemistry Technician...........................
6.3 Repairing Snubber on Emergency Diesel Generator Service Water Pipe..........................................................
6.4 Spurious Actuation of Emergency Safeguards Feature, Emergency Diesel Generator..............................................
7.
Maintenance /Surve111anca (61726, 62703)..............................
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Table of Contents PAGE 8.
Review of Periodic and Special Reports (71707, 90712)................
8.1 P e r i od i c Re p o rt s................................................
8.2 (Closed) LER 89-006: Mispositioning of unborated Water Source Locked Va1ves.................................................
8.3 (Closed) LER 89-007: ESF Actuation, Containment Ventilation Isolation................................................
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8.4 (Closed) LER 89-008: Manual Reactor Trip During Natural Circulation Test..............................................
9.
Allegation Followup (92701)..........................................
9.1 Pittsburgh Testing Laboratory Services (RI-87-A-116)............
9.2 Hammel Dahl EFW System Isolation Valves (NRR-89-A-0024).........
10. Management Meetings (30703)..........................................
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DETAILS 1.
Persons Contacted E. Brown, Engineering Supervisor J. DeLoach, Executive Director of Engineering and Licensing -
B. Drawbridge, Executive Director of Nuclear. Production T. Feigenbaum, Senior Vice President and Chief Operating Officer
- D. Moody, Station Manager J. Peschel, Regulatory Services Manager
- N. Pillsbury, Director of _ Quality Programs
- J. Warnock, Nuclear Quality Manager
- Attended exit meeting conducted August 17, 1989.
Other licensee and contractor personnel were also contacted.
2.
Summary of NRC and Facility Activities 2.1 Resident Inspector Activities One senior resident inspector was assigned to the site during the entire inspection period. The inspection included 47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br /> during backshift periods and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> during deep backshift periods.
(Deep backshift inspection was conducted from 8:00 a.m. to 4.45 p.m. on Jul/ 22, 23 and August 12, 1989, and from 8:00 a.m. to 2:00 p.m. on August 13,1989.)
2.2 Visiting Inspector and Management Activities Beginning July 26, 1989, an inspector from the Office of Nuclear Reactor Regulation vendor branch conducted a two day inspection. The results are discussed in paragraph 9.2.
On July 27, 1989, the Region I Projects Branch Chief toured the site and held discussions with senior licensee managers and control room operators.
On August 16, 1989, the Director of the Office of Nuclear Reactor Regulation toured both Unit 1 and Unit 2 and held discussions with senior licensee managers and control room operators.
2.3 Plant Activities The plant remained in operational mode 5, cold shutdown, with primary coolant temperature between 120F to 140F and the reactor vessel vented. Major maintenance was conducted on residual heat removal system train A, the control building air supply system, the primary component cooling water system, and diesel generator A.
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2.4 Facility Management Reorganization
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On June 29, 1989, the Vice President - Nuclear Production was re-lieved of all responsibilities related to Seabrook Station and re-signed. This action. occurred after the June 22, 1989 manual reactor trip event.
On July 19, 1989, New Hampshire Yankee (NHY) announced the following organizational changes which became effective July 24, 1989.
Mr. T. Feigenbaum was elected Senior Vice President and Chief
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Operating Officer of NHY (Division of Public Service of New Hampshire) and assumed responsibility for all. operations at Seabrook. Mr. Feigenbaum reports to Mr. Edward Brown, President and Chief Executive.0fficer of NHY, who remains on site.
Mr. J. DeLoach became. Executive Director of Engineering and -
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Licensing, and reports to Mr. Feigenbaum.
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Mr. N.."illsbury became Director of Quality Programs and reports to Mr. Feigenbaum.
Mr. B. Drawbridge became Executive Director of Nuclear Produc-
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tion, and reports to Mr. Feigenbaum.
Mr. Drawbridge, who was Vice President of' Yankee Atomic Electric Company (YAEC), is to be responsible for Nuclear Production until a permanent nuclear production executive is named.
3.
Status of Previous Inspection Findings 3.1 (Closed) Unresolved Item 87-16-02:
RHR Line Weld Failures Two cf the five open subitems were closed in NRC Inspection Report 50-443/88-06.
Remaining open subitems (2) and (3) dealt with accept-
-ability of specific system lineups with regard tc operability and design bases. Although no violation was identified, it was requested that the licensee review Information Notices 83-56, " Operability of Required Auxiliary Equipment," and 84-42, " Equipment Availability for Conditions During Outages Not Covered by Technical Specifications,"
to assure that adequate guidance is available to control room opera-tors when similar conditions are experienced. The licensee recog-nized deficiencies in their review and dispositioning of Information Notices in general and for review of Information Notices 83-56 and 84-42 in particular.. The licensee completed an audit of the final disposition of Information Notices for the last two years and planned to establish a schedule for completion of Inspection Notices 83-56 and 84-42 by September 1, 1989. The licensee has demonstrated suf-ficient progress in resolving problems with reviewing and implement-ing Inspection Notices that these two subitems are considered closed.
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I The last open subitem (4) concerned the planning and preparation for similar repairs. Through review of operating logs, a miner modifica-f tion design package, and a work request, the inspector noted that, on l
April 13, 1989, with the plant in operational mode 5, cold shutdown, a pin hole leak was identified in the casing drain of charging pump A.
Charging pump B was lined up and became the operable charging-pump. Charging pump A was declared inoperable, isolated, and tagged out until the weld was acceptably repaired.
The inspector concluded that adequate modification planning and preparation have been shown.
Unresolved item 87-16-02 is therefore closed.
3.2 (Closed) Violation 89-03-02: Boron Dilution Flowpath Vaives Found Mispositioned On April 7 and 17, 1989, valves associated with possible boron dilu-tion paths were found mispositioned or unlocked. The licensee issued licensee event report (LER) No.89-006 on April 28, 1989, and re-sponded to the violation in a letter dated June 26, 1989. The in-spector reviewed the LER an.1 the licensee response to the violation, taking inte account the collective significance of these events, and verified aspects of the long and short-term corrective actions in-cluding valve positioning, discussions of the event between shift supervisors and operating t.rews, and procedural changes.
The root causes were identified by the licensee as procedure inade-quacies and inattention to detail.
Short-term corrective action in-volved properly repositioning and locking the mispositioned valve.
After the second occurrence, Procedure 05-86-1-7, "Unborated Water Source Locked Valve List," was conducted to verify that all valves on the list were in their proper position and locked. Cautions and pre-cautions were added to procedure steps where valves were being mani-pulated for operation of letdown, charging, seal injection, excess letdown, and letdown degasifier systems.
The cautions alerted opera-tors to subsequent required valve restoration steps.
Steps were added to surveillance procedure 05-86-1-7 to visually verify the position of capped valves each time the surveillance is performed, track the position of manipulated valves in the " Locked Valve Log,"
and provide determination of required valve positions by a senior reactor operator. Additional training is scheduled for licensed and non-licensed operators on the responsibilities associated with the manipulation and verification of components which have special guid-ance contained on caution tags. The completion of training is being tracked on the licensee Integrated Commitments Tracking System (ICTS).
The inspector considered the collective significance of the April 7 and 17, 1989 events and concluded that the events did not represent a programmatic breakdown. The requirement for locking the boron dilution path valve was established in October 1986 and, for over two years prior to April 1989, the valves were properly controlled. The
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inspector attributed.the failure to adequately control the boron path
valves to operator inattention to detail. Also, the inspector con-cluded that'the significance of the event was that the inattention to
~ detail was not a singular error but occurred over several-shifts and on two separate occasions.
The Vice-President - Nuclear Production issued a memorandum to Pro-duction Managers on May 5, 1989, emphasizing the need to pay atten-tion to detail. Each Shift Superintendent discussed the events with-shift personnel and emphasized the importance of special license-re-quiremants, stressed the requirement for operators to be aware of procedural changes, and stressed the need to pay attention to proce-dural detail.
As a result of licensee ' review of the procedural change mechanism, Station Management Manual Chapter 6.2, " Station Operating Proce-dures," 'was revised on July 17, 1989 to require a change to a proce-dure to be incorporated into the body of the procedure and not at-tached as a separate sheet to the front of a procedure, as had pre-viously been required. Additional training of licensed and non -
licensed operators on the performance of procedures which have proce-dure change sheets attached is planned to be completed prior to ex-ceeding 5% power. Completion of that training is being tracked by the licensee on the ICTS.
A review of station operating procedures associated with the control of locked valves in areas other than boron' dilution flowpaths is scheduled by the licensee to be completed by September 1, 1989. The completion of this review is being tracked by the licensee on the ICTS.
The licensee implemented a Human Performance Evaluation System (HPES)
to evaluate personnel errors and establish methods to prevent recur-rence. That system was found by the inspector to be adequate. The l3 HPES was used to evaluate the June 22, 1989, event which occurred
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during the natural circulation test.
The licensee established an on going task team to recommend ways to minimize human errors and improve the accuracy and timeliness of the event evaluation process. The licensee Self-Assessment Team (SAT)
was requested by senior management to review processes associated with station operation to identify areas where events similar to this event could occur. This evaluation is to be included as part of the SAT Phase II report.
The inspector had no further questions.
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3.3 1 Closed)UnresolvedItem 89-06-01:
Release of Liquid Waste Tank On June 2, 1989, the contents of a liquid waste test tank were re-leased with the radiation monitor high alart set point incorrectly set above the required trip set point.
Initially the licensee modif-ied the liquid effluent release permit in chemistry procedure CS0917.01B, to' require the radiation monitor set points to be ver-ified after.the completion of the functional test _of.the discharge valve. The' inspector asked whether the operational discharge pro-cedure had been properly followed and whether there were generic im-plications for other inter-organizational procedures'.
Through discussions with the licensee the inspector learned that operational discharge procedures required verification of the dis-charge valve trip set point prior to conducting the fuctional test.
Procedure ON1018.08, " Waste Test Tank 63B Discharge to Transition
. Structure," was changed to require a verification of the radiation monitor set point as a last step prior to the alignment-of the dis-charge header for overboard discharge. A similar change was made to the discharge procedures for recovery test tanks 58A and 588, waste test tank 63A, and the waste holdup sump. The inspector had no fur-ther questions.
3.4 (Closed) Unresolved Item 89-80-01: Use of Step List to Bypass Normal Review Cycle Two. step lists to transfer resin from the chemical and volume control system demineralizers were written and used prior to plant heat-up for.. initial criticality.
In discussions with the inspector the lic-ensee stated that, upon further review, the step lists which were used were considered temporary changes to existing procedures. Also,
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the licensee concluded that the intent of the existing procedure was not changed inasmuch as resin was transferred and exposure was mini-mized. Due to the low activation levels and the iron content of the resin, an alternate discharge path was.used. The inspector deter-mined that the step lists did not change the intent of the original procedure and that the requirements of technical specification para-graph 6.7.3 for temporary changes were met. The unit shift super-visor and shift superintendent approved the steplists and the step-lists were reviewed by the Station Operations Review Committee (SORC)
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within 14 days of implementation.
The licensee has modified their procedure for implementing steplists, OP 10.7.
The procedure restricts the use of steplists to situations where there is no affect on reactor safety-and where the regulatory requirements for written procedures contained in Regulatory Guide 1.33 and ANSI Standard 18.7 do not apply.
Steplists may not be used for technical specification surveillance testing, for engineered safety feature system alignments, for evolutions which may create unreviewed safety questions, for evolutions which alter reactor core
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reactivity, cr as a substitute for. approved procedures. The shift superintendent determines whether a steplist should be used and whether additional reviews are required by chemistry, health physics and quality assurance personnel. The licensee stated that the pre-sent requirements of OP 10.7 do not allow the use of a steplist for the temporary change to the resin transfer procedure. The inspector reviewed three recent steplists, found them acceptable, and had no further questions.
'3.5 Painting Schedule for Primary Auxiliary Building (PAB) Floors In NRC Inspection. Report No. 50-443/89-05,'the inspector questioned the adequacy of the licensee long. term schedule to repaint the_ floors of the primary auxiliary building. _The painting schedule was re-viewed by an NRC specialist inspector during the Readiness Assessment Team inspection with regards to the possible exposure of painters and the possibility of fixing contamination. The review determined that the schedule was adequate. The inspector had no further questions.
13.6 Four Hour Emergency Lighting In response to an NRC question concerning the operability of the emergency lighting self-contained battery pack units, the licensee performed a complete walkdown of all units. Discrepancies were iden-Lified between field conditions, drawings, the foreign print manual and the bill of material. The walkdown determined that there were more installed emergency lighting units than indicated on the draw-ings and that the control circuits and locations-of some of the units
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had changed. Document Revision Report (DRR) 89-0055, " Emergency l
Lighting Self-Contained Battery Pack Units and Protected Area," was
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issued on July 6, 1989, to correct the identified discrepancies on the prints and drawing. No field work was necessary. The inspector reviewed the DRR and had no questions.
3.7 Nuclear Safety Audit and Review Committee In NRC inspection report 50-443/89-03, the inspector noted that an appar-ent excessive volume of material was required to be reviewed by the Nuc-lear Safety Audit and Review Committee (NSARC).
In response to this concern, a letter dated July 18, 1989 was written from the Nuclear Quality Manager to the Vice President, Engineering and Quality Pro-grams, to document changes to the reviews required to be conducted by the NSARC. The inspector reviewed that letter and determined that subcommittees have been established to review and report on security problems, design change safety evaluations, and Station Operation Review Committee minutes. Audit reports and NRC Inspection Reports are now summarized and reported to the committee by single indivi-duals. The station information reports are now reviewed and screened by the Independent safety Engineering Group prior to review by the NSARC. The inspector had no further questions.
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3.8 Service Water Supply Valve SW-18 Failure
. The failure and repair of service water supply valve _ SW-18 and-the subsequent modification of service water supply SW-16 are described-
- in NRC Inspection Report 50-443/89-03. The licensee submitted a 30-day special report. No other valves in the plant were identified.as -
subject.to a similar failure. The drive bushing on SW-18_is sched-uled for inspection in October 1989 to verify that the bushing has-not slid down the valve shaft. As'an enhancement to.SW-18, a poly-ethelene spacer is planned for installation below the bushing when o
the actuator is removed from.the system for maintenance or inspec-
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tion. A modification of the seismic supports associated with SW-16
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and 18 was initiated to allow easier removal of the valve actuators
for maintenance. The inspector had no further questions.
3.9 Main Steam-Isolation Valve Actuators Rockwell International Corporation, the manufacturer of the Type A
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main steam isolation valve (MSIV) actuators used at Seabrock, noti-fied the NRC by letter dated January 23, 1989 of a potential for.the MSIV actuator pistons developing a debonding of their aluminum bronze coating.
Rockwell stated that they did not consider that the poten-tial debonding represented a substantial safety hazard. The,11cen-see's_ initial response to this concern is discussed in NRC Inspection Report 50-443/89-01.
Based on further discussions with Rockwell and a letter received ori February 14, 1989, the licensee plans to repair the MSIV pistons.
Through review of-meeting minutes and discussion with the licensee, the inspector learned that the four Unit 2 MSIV actuators will be sent to Rockwell for modification and will be installed in the Unit 1 MSIVs during the first refueling outage, currently scheduled for 1991. The inspector had no further questions.
4.
Operational Safety 4.1 Plant Operations Review The inspector observed plant operations during regular and backshift tours of the control room, prim;ry auxiliary building, containment, service water building, service water cooling tower, diesel building, and protected area fence line. Control room instruments were ob-served for proper functioning and conformance with technical speci-fications. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators. Opera-tors were found cognizant of board and plant conditions. The inspec-tor reviewed the tag-out log and identified several discrepancies in the index. These were resolved by the licensee. The inspector ac-companied a fire watch on his hourly tour.
No unacceptable condi-tions were identified.
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4.2 ' Use of Unqualified Tape 'in Primary Auxiliary Building During a tour of the primary auxiliary building (PAB), the inspector questioned the qualification lof Nashua. Nuclear. Duct Tape, which was-being used to_ establish inerted conditions for butt welds on stain-less steel. piping.. Quality Assurance Surveillance Report (QSAR) 89-
~ 00656 was immediately initiated. Nashua tape was removed from the PAB and prohibited from being issued from the store room.
Nashua tape is planned to be removed.from site.
The inspector reviewed the response to the QSAR and learned that only
. Tuck Tape Industry. tape.had been approved and qualified for use. A Certification of Compliance was obtained by the licensee from Nashua Corporation and a. chemistry evaluation was conducted on a representa-
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tive. sample of Nashua tape. The chemistry evaluations identified leachable chlorides above the limits contained in.the Certificate of.
Compliance. The licensee verified approved cleaning techniques by cleaning and swiping stainless steel piping which had been removed from the field and capped off using Nashua tape. Subsequently, the piping areas where the Nashua tape had been used were cleaned and swiped to verify the removal of-any deposited halogens.
A request for engineering support was generated by the licensee to analyze the effects of the use of Nashua tape on the quality of welds which had been completed.
The inspector concluded that the licensee should have recognized the loss of control of Nashua tape sooner.
However, the licensee fol-low-up actions to the inspector's concerns were well directed and thorough, and the inspector had no further questions.
5.
Security 5.1 Review of Internal Security Incident Reports The inspector reviewed approximately 30 licensee internal security incident / complaint reports written between January 1,1987 and July 1, 1989 about possible contraband in the protected area. The deportability of the events was reviewed based on Regulatory Guide 5.6.2, which provides guidance for irrlementing 10 CFR 73.71, which became effective on October 8, 1987. One formal report made to the NRC during the period reviewed concerned contraband found in the protected area and described in NRC Inspection Report 50-443/89-03.
Of the approximately thirteen internal reports written between January I and October 8,1987, three incidents involving identifi-cation of marijuana or drug paraphernalia would have been considered for deportability had the rule been effective. The inspector con-cluded that such events had been properly dispositioned by the lic-ensee and had no further questions.
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5.2 Transfer of Special Nuclear Material
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On July 11, 1989, the inspector was informed that the return of a
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nuclear detector containing special nuclear material from Gamma Metrics Inc., following repair of the instrument, did not include the Form 741 which is required by 10 CFR 70.54 for transfer of special nuclear material. The licensee contacted Gamma Metrics and informed Martin Marrietta, the NRC contractor with oversite responsibility for l
transfers of special nuclear material, that the proper transfer form had not been received. Gamma Metrics, which had properly documented previous transferral, provided a properly completed Form 741 to the licensee on July 17, 1989. The inspector reviewed the Form 741 and had no further questions.
6.
Plant Operations 6.1 Unmonitored Release from the Turbine Building Sump On July 9,1989, at approxi.-ately 8:20 p m., the temparary sump pump installed in the turbine building sump was started by the auxiliary operator to pump water from the sump to the chemical treatment pond.
Twenty minutes later, a second operator noted that the pump was run-ning, that no chemistry sample had been taken, and that the action statement of technical specification 3.3.3.9 had not been entered.
Chemistry was called, the sump was samoled, and technical specifica-tion 3.3.3.9 was entered. The temporary sump pump was removed from the sump ana the switch to tie pump was cautioned tagged to require notification of the main control room prior to using the pump.
The temporary sump pump provided an unmonitored release path to the environment. That path was not directly addressed by the technical specifications or by an operatirg procedure.
The inspector ques-tioned whether the installation of the temporary sump pump met design criteria, whether adequate controls over releases had been estab-lished, and whether the intent of the technical specifications had been met. Also, the inspector questioned the generic implications of the installation and control of temporary sump pumps used in other areas of the facility. Based on concern for the adequacy of the de-sign review of the temporary modification and with the compliance with technical specification effluent monitoring requirements, this issue is unresolved (UNR 50-443/89-08-01).
6.2 Investigation of Chemistry Technician On July 14, 1989, the inspector was informed that a chemistry techni-cian had resigned following a licensee investigation of his perform-ance of a technical specification surveillance the previous day. The inspector reviewed the licensee's comprehensive description of the licensee's concerns and the actions taken to assure proper completion of technical specification surveillance requirements.
The inspector
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concluded that technical specification-surveillance. requirements for the wide range gas monitor and. chemistry sampling had been properly performed and that there was no safety concern connected with the licensee's investigation findings. The inspector had no further questions.
6.3. Repair of Snubber on Emergency Diesel Generator Service Water Pipe i
The inspector reviewed both the initial Station Information Report
(89-049) and the Quality Assurance Inspection Checklist / Report con-cerning the uncontrolled removal of a snubber from emergency diesel generator (DG) B.
The inspector determined that, on August 3, 1989,
a snubber was removed from the DG-A water jacket heat exchanger pip-i ing in accordance with a work request and that technical specifica-tion (TS) 3.7.7, which requires snubbers to be operable, was not entered since DG-A was not required to be operable. On August 4, 1989, the system engineer changed the scope of the work request to include the snubber on DG-B and informed quality control of the change.
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On August 4, the main contrel room approved the change of scope of l
the work request without recognizing that DG-B would be affected. On i
August 5, a licensee quality control inspector recognized that DG-B
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was affected by the removed snubber and informed the main control
room. The shift superintendent entered TS 3.7.7, which required the
snubber to be reinstalled within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The snubber for DG-B water jacket was re nstalled within the required time limit. TS l
3.7.7 was exited on August 5.
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Through discussions with the licensee staff, the inspector learned that several steps are being taken to improve the centrol of snubber removal. As a result of a previous Station Information Report, the j
licensee developed a list of snubbers which affect safety-related
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equipment but decided that it was impractical to develop a list of all safety-related equipment which was affected by each snubber due to the various possible plant modes and system configurations. The licensee has concluded that, to preclude further problems, whenever a i
snubber is removed and is on the list of safety-related snubbers, TS 3.7.7 will be entered.
If the snubber is not reinstalled in 72
hours, an engineering evaluation is to be required to determine sys-
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tem operability at the end of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The licensee is pre-i sently writing, reviewing and approving a procedure which implements the above philosophy. The inspector concluded that these steps will be adequate to control the removal of snubbers and had no further questions.
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'6.4 Spurious Actuation of Emergency Safeguards Feature, Emergency Diesel Generator
On August 15, 1989, the inspector was notified by the licensee that
. an Emergency Notification System call had been made when emergency
. diesel. generator (DG) B was automatically started and loaded due to electrical bus 6 being deenergized. The loss of power to bus.6 was due to a personnel error which occurred whi?e clearing tags when the
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wrong drawer in the breaker cubical was opened and the potential I
I transformer fuses were disconnected on bus 6.
Power was restored to
. bus 6 from the normal off-site supply, and DG-B was secured at ap-l L
proximately 7:53 p.m.
Pending receipt'of the licensee's event re-
pcrt, the inspector had no further questions.
7.
Maintenance / Surveillance The inspector tracked several major maintenance activities which were in progress throughout the inspection period. The tracking of the activities
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included observation of selected activities and review of selected work requests and procedures. A listing of the work requests and procedures reviewed is provided belr.w.
DCR 87-422 Replace Residual Heat Removal (RHR) pump miniflow l
valves 1-RH-FCV-610.
89W002210 RHR Recirc Valve RM-303, RHR train A.
89W00173 Hot leg isolation V-29: replace limitorque motor pinion key with new key.
DCR 87-311 Check valves installed in RHR pump suction lines..
89 WOO 3244 1-RH-V-31, RH-A, CI-205 check valve leakage.
MS0526.07 Freeze Seal of Piping.
89 WOO 3180 RH-E-9A A Train RHR heat exchanger head gasket leaking.
89W001510 10CC-FISHL22248/2147;. replacement of inline Ray-chem splice for three leads.
ES87-1-12 Raychem Splice Inspection Program, Revision 1, Change 1.
MS0514.09 Low Voltage Raychem Installation, Revision 1, Change 4.
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.MX0507.03 Thermal Overload Relay Replacement for Motor-Operated Valves.
89RM1732001 Thermal Overload Relay Replacement.
MX0507.05 Calibration of Thermal Overload Protection Relays.
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The maintenance and surveillance activities were found to'be well con-
. trolled and supervised.- Work was controlled through the main control room
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and adequate communications were conducted to assure sufficient tag outs were established prior to the commencement of any job. Health physics personne1'were present for jobs which breached primary system boundaries.
Quality control personnel were present at appropriate points during main-tenance and surveillance activities, and raised appropriate. issues when questionable work practices were noted. The technicians followed the directions provided in applicable work requests and procedure.
Problems encountered while performing procedures were' discussed and resolved with supervisory personnel who were routinely in the work area.
After the inspector' questioned the adequacy of a t:mporary. ASME Level II-storage' t.rea and the general housekeeping of a particular area, improve-ments in storage and housekeeping were noted.
The inspector concluded that maintenance and surveillance were being ade-quately performed and that the organization was respoasive to NRC con-cerns.
8.
Review of Periodic and Special Reports 8.1 Periodic Reports The inspector reviewed the following periodic reports and had no questions.
Monthly Operating Report 89-06 issued July 13, 1989
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Monthly Operating Report 89-07 issued August 9,1989
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8.2 (Closed) Licensee Event Report (LER) No. 89-006: Mispositioning of
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Unborated Water Source Locked Valves l
Refer to NRC Inspection Report No. 50-443/89-03 and paragraph 3 of this report for discussion and closure of this item.
8.3 (Closed) Licensee Event Report (LER) No. 89-007: Engineered Safety Feature Actuation - Containment Ventilation Isolation This event is described in NRC Inspection Report 50-443/89-06. The inspector reviewed the LER and determined that the root cause of the event was a defective Geiger-Muller (GM) tube. The GM tube was re-
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placed and the system was tested and declared operable. The licen-see's long-term corrective action was to review data developed during calibration of radiation monitors to determine if a method can be developed to predict GM tube failures. The inspector had no ques-tions.
8.4 (Closed) Licensee Event Report (LER) No.89-060 Manual Reactor Trip During Natural Circulation Test An evaluation of the June 22, 1989, natural circulation test event is documented in NRC Inspection Report 50-443/89-62. A Confirmatory Action Letter was issued by the NRC on June 23, 1989, confirming the understanding-that the licensee would determine the root causes of the event, establish short and long-term corrective actions and re-view the results with the NRC prior to restarting the unit.
In ad-dition, a meeting was scheduled for September 7, 1989 to discuss the regulatory significance connected with the June 22 event.
For administrative purposes, this LER is closed. Adequacy of licen-see actions will be addressed incident to processing of Report 50-443/89-82.
9.
Allegation Followup 9.1 Pittsburgh Testing Laboratory Services (RI-87-A-116)
l In October 1987, the NRC was informed of a concern about the perform-l ance of Pittsburgh Testing Laboratory (PTL) personnel at Seabrook Station.
PTL, a contractor during the construction phase of Seabrook Unit 1, provided ce'*truction material testing and inspection ser-l vices at the site, oarily in the area of structural concrete and L
rebar testing and 1.r ction.
The PTL contract for work at Seabrook Station ended in March 1986.
The allegation was general and related to fitness-for-duty of a PTL employee or employees.
No direct allegations of improper construc-tion, quality or work were raised. The Region I Administrator trans-mitted a letter to the licensee on March 18, 1988, requesting re-sponses to questions dealing with the Seabrook fitness-for-duty pro-gram. Two of these questions dealt specifically with PTL performance and the related assurance of work quality.
The licensee response dated May 16, 1988, addressed the specific questions regarding drug incidents, the termination of the PTL con-tract, and evidence of assurance of site concrete and construction quality at Seabrook Station. The responses relating to PTL were re-viewed by the senior resident inspector and by a regional specialist inspector.
Licensee personnel were interviewed and records were re-viewed to assess the circumstances and the chronology of the inci-dents which served as the basis for the alleger's concerns. This
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inspection and review effort identified no impact on the quality of structures and materials due to PTL testing and inspection services.
The inspectors identified no evidence of a weakness in the Seabrook project's quality assurance program that could lead to a failure to detect deficiencies in the quality of the plant's safety-related structures, systems, components and materials, and.no evidence that the general fitness-for-duty concerns raised by this allegation had any adverse safety impact.
In addition, the results of system and
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I component acceptance testing, startup testing, operational testing,.
and surveillance testing Conducted after the Construction phase do not support an allegation that drug usage adversely impacted con-struction, in that the test results have shown acceptable equipment and structural conditions.
The allegation was not substantiated; no further inspection is planned.
9.2 Hammel Dahl Emergency Feedwater System-Isolation Valves (NRR-89-A-0024)
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During the current inspection, the inspector reviewed an allegation about Hammel Dahl (HD) emergency feedwater system isol:. tion valves FW-393, 394 and 395. The allegation was that the isolation valves were counterfeit and questioned the original source of manufacture of the subject valves, the original rework facility and the source of the internal valve replacement parts (valve trim)
Tne inspector reviewed records associated with the four (including one spare) HD L
valves. Those records date back to 1978.
United Engineers & Con-structors, Incorporated (UE&C) originally procured the subject valves
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from ITT Hammel Dahl, 1175 Post Road, Warwick, Rhode Island, for use i
at the Washington Public Power Supply System's (WPPSS) Washington l
Nuclear Plant Unit No.4 (WNP-4), under WPPSS specification 9779-42, I.
in 1980. Subsequently, WNP-4 returned the valves to HD and the l-valves were then procured again by UE&C, in 1986, this time, for the i
Seabrook Station. The record review revealed that the valves were I
cast by the Quaker Alloy foundry in Pennsylvania, in approximately 1979, and were uniquely identified by HD as required by WNP-4 as valves: (1) FWA-LCV-4007; (2) FWA-LCV-4009; (3) FWA-LCV-4025; and (4) FWA-LCV-4026.
The inspector identified the applicable foundry heat code numbers, valve serial numbers and other unique markings from the UE&C/WPN-4 documentation for each of the four valves listed above. The valve and bonnet identification numbers were traced back through the dif-ferent tiers of documentation in the four packages. The inspector verified that traceability was maintained and can be substantiated beginning from the original casting of the valve bodies at the foundry, through two UE&C procurement processes from Hammel Dahl, and ultimately to the Seabrook installed hardware.
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The Seabrook valve assembly procurement document packages contained documents such as: (1) ASME NPV-1 Certificate Holders Report for Valves; (2) HD valve data record binder reassignment forms; (3) HD certificates of compliance; (4) coating certificates / qualifications; (5) paint reports; (6) HD QA assembly and functional checklists; (7) HD quality records index for UE&C p0 9779-42; (8) HD QA component part checklists; (9) Quaker Alloy Casting Company (Quaker) material test reports; (10) Quaker certificate of acceptance on quality exam-ination and inspections; (11) Quaker certificates of liquid penetrant examination (PT) personnel; (12) Quaker assorted casting, welding, qualification and procurement documents; (13) HD valve drawings; (14) HD PT records; (15) HD test reports; (16) HD material certifi-cates; (17) Carpenter Technology test certificates; (18) Teledyne-
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McKay material test reports; (19) Texas Bolt Company certificates of test; and (20) other associated documents.
An additional verification effort was performed to determine whether the installed hardware was as purported by the documents. The three installed HD valves and the spare valve body and bonnet stored in the Scabrook warehouse spare were visually examined.
Each valve was marked with the appropriate serial number, heat code number, and the HD trade mark. All of the serial and heat code numbers found on the valve assemblies, including valve trim, matched t69 corresponding numbers found in the above discussed documents. ine as-cast HD trade mark, Hammel Dahl - ITT, was prominently displayed on one side of each of the valve bodies.
Each of the four valve body serial numbers was also observed as being " cast" onto a raised " pad" on the side of each valve opposite the side bearing the trade mark.
In conclusion, each of the four subject valves are the original equipment manufacturer's product as represented by the records and as procured for nuclear service. The allegation was not substantiated; no further inspection is planned.
10. Management Meeting At periodic intervals during this inspection, meetings were held with senior plant management to discuss the findings. A summary of the report was also discussed at the conclusion of the inspection.
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