IR 05000352/1989009
| ML20247E026 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 05/11/1989 |
| From: | Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20247E013 | List: |
| References | |
| 50-352-89-09, 50-352-89-9, 50-353-89-15, IEB-88-007, IEB-88-7, NUDOCS 8905260098 | |
| Download: ML20247E026 (24) | |
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U.S. NUCLEAR REGULATORY C0i4 MISSION
REGION I
Report No.
50-352/89-09 50-353/89-15 License No. NPF-39 CPPR-107 Licensee:
Philadelphia Electric Company Correspondence Control Desk P.O. Box 7520
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philadelphia, Pa 19101 Facility Name:
Limerick Generating Station, Unit 1 and 2 Inspection Period: March 29 - April 23, 1989 Inspectors:
T. J. Kenny, Senior Resioent Inspector
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L. L. Sch 11, Resident Inspector R.L.F rme
, Resident Inspector
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Reviewed by:.
'H.W ill
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'ect Engineer Date
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Approved by:
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J ' W1Srivi l l e,
ieff, Projects Section 2A Date'
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Summary:.Ru ne daytime (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and backshift/holidcy (45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />) inspec-tions of Unit I and 2 by the resident inspectors consisting of (a) plant tours,
.(b) observations of maintenance and surveillance testing, (c) review of LERs
- l and periodic reports, (d) review of operational events and (e) system walkdowns.
Areas Inspected:
Resident safety inspection of the following areas:
opera-tions, radiological controls, surveillance testing, maintenance, emergency pre-paredness, security, engineering / technical supnort, safety assessment / assurance of quality, Unit 2 preoperational test program, review of licensee event re-ports and open item followup.
Results:
Unit 1.
The inspection period documents inspection during the latter phase of the current. refueling outage.
Several areas of concern were identified by the inspectors.
1) Removal of the packing from the wrong valve.
The licensee per-l formed an indepth study and is currently taking corrective actions (section
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2.1.2).
2) Battery charger surveillance testing indicated a lack of engineer-ing judgement.
Inspector questioning caused the licensee to reassess the test and perform it over (section 4.1.1).
3) Violation of procedure adherance while performing maintenance on Main Steam Safety Valves (section 5.1).
Unit 2.
The inspection documents the review of " Proof and Review" Technical
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Specifications (section 9.0) and the close out of various open items necessary I
to support licensing (section 4.0).
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DETAILS l
1.0 Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspec-tion activity.
2.0 Operational Safety Verification and Plant Tours 2.1 Unit 1 (71707)
2.1.1 The inspector conducted routine entries into the protected areas of the plant, including the control room, reactor enclosure, fuel floor, and drywell (when access is poss-ible).
During the inspection, discussions were held.with operaters, technicians (HP & I&C), mechanics, security per-sonnel, supervisors and plant management. The inspections were conducted in accordance with NRC Inspection Procedure 71707 and affirmed the licensee's commitments and compli-ance with 10 CFR, Technical Specifications, License Con-ditions and Administrative Procedures.
2.1.2 Inspector Comments / Findings The report begins with the unit in a refueling condition.
On March 29,1989, at 11:33 p.m., a refueling floor isola-tion occurred due to a low differential pressure (DP) con-dition. The standby gas treatment system started and operated as designed.
Secondary containment was not re-quired at the time since no core alterations were in pro-gresh.
The cause has been attributed to the decreasing temperature of the outside atmosphere without heating steam in the air supply plenum.
Following the initiation of the steam supply to the heating coils the DP returned to normal and the systems were returned to normal service.
j On April 3,1989, at 2:04 p.m., during testing, a LOCA signal was initiated when switch E214-S22A was placed in the armed position and the push button was depressed by the operator. This action actuated shunt trip relay E21A-K18A which caused the D114 load center breaker to open, however no major equipment started due to the plant configuration at the time of testing.
The licensee identified a faulty procedure that omitted the resetting of the aforementioned relay which would have prevented the load shed sequence to occur.
The licensee restored conditions to normal, cor-rected the procedure and resumed the retest. The licensee informed the NRC via the Emergency Notification System
(ENS).
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i l-t On April 5, 1989, during an engineering review of the Ap-pendix R Analysis, the licensee discovered that no unaf-fected temperature or level display exists at the remote shutdown panel for the suppression pool. This could pre-vent tha suppression pool from performing its safety func-tion. The suppression pool temperature and level instru-ments presently installed at the remote shutdown panel are operable, however, their signal and power cable may not be adequately protected against fire damage. The lfcensee is conducting an ongoing investigation of Appendix R features and will report the results in an LER.
The resident will review the results when the LER is published by the licen-see.
On April 7, 1989, a 3echtel pipefitter was assigned t.o re-pack valve HV-087-1518 (MRF 8902526).
In preparat138 for performing the job the craftsman reviewed the work sckage, which included a work area location drawing, and proceeded to " dress out" so as to gain entry to the drywell. As part of his preparation for entry, the craftsman denuted the abbreviated equipment number "HV151B" on a piece of tape on his arm. The craftsman proceeded to the drywell area that he thought was indicated on the work package location draw-ing.
In this area the craftsman located a "HV151B" (see attachment 1); and proceeded to remove the valve packing.
The valve the craftsman had located and proceeded to work on was not HV-087-151B but rather HV-151-151B, RHR Return Check Valve Bypass Valve. The IB RHR loop, which had been in shutdown cooling mode of operation at the time of the event, was removed from service and the 1A RHR loop was placed in the shutdown cooling mode of operation within one hour to satisfy Technical Specification 3.9.11.1.
The lic-ensee performed a Human Performance Behavioral Analysis (HPBA) and documented the following results.
Written Communications - the location drawing was deficient in information, pertially illegible and too generic.
Equipment Condition - labelling:
valve marking was inade-quate (see attachment 1).
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Work practices i
1.
Self-checking was not applied to ensure correct component for work.
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2.
Partial equioment number (HV151B) was used in preparation of component location.
3.
Not having proper instructions in Drywell (Radwaste consideration) (MRF, location drawing).
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Work Package Planning - The work package location drawings do not have a consistent method of specifying component location.
Training / Qualification Content The content did not adequately address:
1.
The component / equipment identification 2.
References used to perform task 3.
Verification /self-check practices The HPBA study recommended:
1.
If the location drawing is to be included in the work package it should contain clear, complete information regarding the equipment location such as equipment number (as listed in the equipment index), area eleva-tion and Azmith.
It may be advantageous to utilize a stamp on location drawing.
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Equip:
Area:
Elev:
Az:
This should be stated in Work Package Preparation Guideline.
2.
Remove the hand written "HV151B" marking and mark the valve properly with "HV-151-151B" on actuator.
3.
Craft indoctrination should emphasize unique equipment ID tags and their use in component identification.
4.
Craft indoctrination (craftsman training) should address use of full equipment numbers at all times.
5.
For jobs which are performed in Radioactive Work Per-I mit (RWP) controlled areas the work package should be located at the work area stepoff pad.
Additionally, a copy of the work package and applicable location or isometric drawings should be brought into the RWP area to facilitate equipment identification.
6.
Work package indoctrination for all craft personnel should address component identification.
Including:
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on work package and as shown on equipment'ID a -
tags which includes system numbers.
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discussion of self-checking practices to verify l
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correct component.
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location drawings blocking permits Licensee management is currently considering the recommendations for incorporation into the maintenance program.
The resident inspectors will follow the licensee's progress.
On April 9, 1989, at 10:31 a.m., while performing l
surveillance on the drywell pressure switches a fuse blew causing a shutdown cooling isolation.
The fuse blew when the technician connected a recorder, as directed by procedures, into the system to measure time responses of the pressure switches.
Unknown, and subsequently discovered, was an internal short in the recorder that caused a spike on the control system resulting in the blown fuse. The licensee replaced the fuse and returned the system to normal at 11:11 a.m.
The temperature of the primary system increased by two degrees over the 40 minute period. The surveillance was subsequently completed:
without incident. The licensee informed the NRC via the ENS.
On April I?. 1989, at 1:24 a.m., an unauthorized contractor employee entered the control room as another worker was exiting.
The contractor employee did not have authorization to access the main control aoom and claimed l
he did not realize he was entering the control room.
The j
unauthori7ed entry was discovered when he attempted to i
exit the area.
Subsequent licensee investigation determined that the individual was authorized for access to the control room but access had not been provided.
The licensee informed the NRC via the ENS.
On April 21, 1989, the licensee notified the NRC via the ENS that a contractor employee had introduced a controlled substance into the protected area.
The employee's access has been terminated in accordance with the fitness for duty program.
The licensee is continuing to investigate the incident and is keeping the resident inspectors appraised of their findings.
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On April 22, 1989, the dimension 2,000, the Prelude Tele-phone system and ENS communication system were out of ser-vice when the telephone transformer power supply was de-energized in order to transfer from temporary power supply back to the normal power supply. A temporary feed had been utilized for outage testing.
The ENS system was restored to service within 30 minutes.
2.2 Unit 2 2.2,1 Plant Inspection Tours (337301, 370329, 370316, 370315, 370312)
The inspector witnessed portions of the flow balance on the
"A" and "C" Loop of the Emergency Service Water (ESW) Sys-tem.
During head-capacity curve verification for the
"C" ESW pump the low discharge pressure alarm annunciated at a flow of 6,000 gpm.
The head-capacity data taken for the flow balance, when plotted on the IST curves, fell outside the acceptable range.
This condition has.been referred to pECo nuclear engineering for review and evaluation.
The resident inspector will review the results of this evalu-ation when completed.
The inspector witnessed portions of the loss of offsite power preoperational test (2P-100.1).
The test was well controlled and conducted in accordance with the procedure.
Test coordination was good with advanced warning of initi-ations over the public address system in order that field personnel would know to stand clear of equipment and cables, and manually prelubricate the emergency diesel engines to minimize test-related stresses and wear.
Start-up management personnel and quality organization personnel were in the control room monitoring the test.
Several test exceptions' were written as a result of an apparent calibra-tion problem with time-delay undervoltage relays.
Resolu-tion of these test exceptions will be reviewed with the test results.
3.0 Update of Open Items 3.1 Unit 1 a.
(Closed) Unresolved Item No. 50-352/88-07-02.
Relating to docu-mentation of stem cleaning / lubrication of air operated ESW valves that exceed stroke time alert limits.
The inspector verified that the licensee's procedures ST-6-011-203-0 Rev. 2, ST-6-011-206-0 Rev. 3, ST-6-011-203-1 Rev. O and ST-6-011-206-1 Rev. 3, have been revised to provide communication feedback of l
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stem cleaning / lubrication. Specific procedural steps have been modified to require documentation of stem cleaning / lubrication.
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This unresolved item is closed.
3.2 Unit 2 (392701)
a.
(Closed) Construction Deficiency Report 50-353/84-00-19.
Mal-functioning ASCO Solenoid Valves.
This item identified the in-appropriate use of four way solenoid valves on air actuated valves in the ESW system.
The inspector reviewed the following documents:
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PECo Letter, Kemper to Murley, dated September 5,1984 (SDR 157)
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PECo Letter, Kowalski to U.S. NRC, dated September 1,1987 (SDR 157)
Nonconformance Report (NCR) 10188, Rev. 15
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The solencid valves for the air actuators were replaced with new Class 1E qualified solenoid valves.
This item is closed.
b.
(Closed) Bulletin (50-353/79-BU-02).
This bulletin identified inadequacies in base plate bolting related to flexibility of the base plates. The inspector reviewed the following documenta-tion:
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PECO Letter, Boyer to Grier, dated July 6,1979
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PECO Letter, Boyer to Grier, dated January 7,1980
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Specification C-64, Installation of Expansion Type Anchor Studs and Shells
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Drawing No. 8031-C-615, Rev. 24, Standard Anchor Bolts Schedule and Details The Limerick program for anchor bolts was approved in NRC In-spection Report 50-352/84-55. The inspector verified that the program requirements were carried over to Unit 2.
This item is closed.
c.
(Closed) Bulletin (353/75-8U-03).
Incorrect lower disc springs and clearance dimensions in series 8300 and 8302 ASCO solenoid valves. This bulletin identified problems caused by improper parts and or clearances in ASCO solenoid operated air pilot j
valves. The inspector reviewed the following documentation:
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PECo Letter, Boyer to O'Reilly, dated May 14, 1975
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NCR 10188, Rev. 15 ASCO Letter, Brown to McDaniel, dated December 2,1983
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The inspector determined that ASCO soleno'd valves used in safety-related applications were replaceo with valves redesigned to eliminate the problems noted in the bulletin.
This item is closed.
3.3 The resident inspector conducted an assessment of Temporary Instruc-tions (tis) at the Limerick Station.
The following table is the cur-rent status of regional and NRC temporary instructions (tis).
Il Title Status Report No.
2500/17 Inspection Guidance for Heat Shrink Open
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Tubing 2500/20 Implementation of ATWS Rule 10CFR50.62 Open
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2500/22 Collection of Collated TLD Measure-Open
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ment Results 2500/26 Bulletin 87-02 Fastener Testing Closed 50-353/89-11
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2515/73 MOV Common Mode Failures Due to Open
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Improper Switch Settings i
2515/82 Minimum Flow Logic Problems that Closed 50-352/86-06
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could Disable RHR Pumps 2515/87 Implement Reg. Guide 1.97 Improved Closed 50-352/89-01 Accident Monitoring 2515/89 BWR-Stress Corrosion Cracking of Open
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Stainless Steel Piping 2515/90 BWR-Scram Discharge Volume System Closed 50-352/87-13 Capability 50-352/87-28 2515/92 Team Inspections of Emergency Closed 50-352/87-08 Operating Procedures 2515/93 Verification of QA Request Regarding Open
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Diesel Generator Fuel Oil 2515/95 Verification of BWR Recirc. Pump Trip Open
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2515/97 Maintenance Team Inspection Closed 50-352/89-80 2515/99 Implementation of Requested Actions Open
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to Prevent BWR Power Oscillations 2515/100 Receipt, Handling and Storage of Closed 50-352/87-24 Diesel Generator Fuel Oil RI 86-01 Inspection of Standby Gas Treatment Closed N/A Limerick System RI 86-02 Inspection of GE AK-F-2 Breakers Closed N/A Limerick RI 86-03 Inspection of GE HGA Relays Open
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RI 87-03 Storage of Transient Equipment Closed 50-352/86-09 50-352/87-21 RI 87-04 Diesel Generator Trips Closed 50-352/87-24 RI 87-06 Diesel Ganerator Air Start Motors Closed 50-352/88-10 50-352/89-04 RI 87-07 Batteries Partial 50-352/88-13 Closed
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i 3.4 Respons'e to NRC'Bulletin 88-07 Supplement 1 " Power Oscillations in Boiling Water Reactors"
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On April 3 and 4, ~ 1989 the inspector reviewed facility documents as indicated in' attachment 2 to the inspection report, interviewed three Shift Technical Advisors (STA), three reactor operators, and two'
sentar reactor operators to determine their knowledge of the proce-i dures and the event at LaSalle th:" initiated the bulletin. The in-
spectors. also interviewed the reactor engineer and a training in-
.j structor.
J The inspector concluded that the operators and STAS are aware of the
. power oscillation event that occurred at LaSalle. The inspector also
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concluded that the operators and STAS were aware of monitoring power
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oscillations by use of the APRMs but were less aware of the use of
.i LPRMs to monitor power oscillations. Due to the procedure concerns I
identified below, additional training will be provided which will enhance the LPRM determinations of power oscillations.
The inspector determi.'ed that the licensee had not fully implemented
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the bulletin in two areas. The bulletin defines. evidence of ' power oscillations as "APRM peak to peak oscillations o' greater that 10%
or periodic LPRM upscale or downscale alarms.in u dition to the guid-ance provided in SIL-380, Revision 1". The facility procedures ad-dressed individual LPRM oscillations of 10% but did not include the
periodic LPRM upscale or downscale alarms as evidence of power oscil-
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lations. The licensee agreed to revise the procedures to include LPRM l
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upscale or downscale alarms as evidence of power oscillations. The L
bulletin also indicated that any time the plant enters the reg;on of
potential power oscillation, and power oscillations are evident, the operator shall manually scram the reactor. The inspector identified that several procedures GP-2, GP-3, GP-4, OT-100, OT-102, OT-103, ON-114, and Appendix C of RE-201.have the potential to enter the region of potential instability and did not have the. instruction.to
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manually scram the reactor if power oscillations were determined. The
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licensee agreed to review and revise the procedures to assure that
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the required instructions are included in the appropriate procedures.
The licensee also agreed to provide training to the operators after i
procedure modifications were made.
J The inspector also noted that the OT procedures utilized different
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wording of instructions on how to reduce power during a transient condition. The licensee representative indicated the licensee is in the process of revising the OT procedures to reference the reactor engineering shutdown instructions on the proper method to reduce
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reactor power when required during the use of transient procedures.
This will reduce the potential for operators to enter the region of potential oscillation.
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g Based on the inspector findings and the actions required of the lic-ensee to fully implement the bulletin, the bulletin is considered open.
See Attachment 2 for a list of documents reviewed.
3.5 Update of TMI Action Items (Unit 2 Only) (425401)
a.
(Closed) II.B.4.
Training for mitigating core damage. The inspector reviewed the following documentation:
FSAR Sections 13.2.1.4, 13.2.2.1 and Quostion 630.17, 1.13
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Station Procedure A-50, Procedure for Conduct of Station
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Training
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Memo, Firth to MacAinsh, dated April 13, 1989 The appendices to procedure A-50 delineate the training required for station personnel.
The inspector verified that core damage mitigation training was included for operations personnel and that I&C, HP and chemistry personnel are trained on emergency response duties and procedures.
This item is' closed, b.
{ Closed)I.D.1.
Control room design reviews (CRDR). The inspector reviewed the following documents:
PECo Letter, Kemper to Murley, dated October 27, 1988
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Finding Report 2E-520 Finding Report 2E-521
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The human factors improvements identified in the Unit 1 CRDR have now been implemented for Unit 2.
The Unit 1 and 2 control panels are now essentially identical.
This item is closed.
c.
(Closed) I.G.I.
Training during low power testing. The I
inspector reviewed the following documentation:
Limerick Safety Evaluation Report (SER) Section 34 I
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Station Procedure E-1, Revision 6, Loss of all AC Power
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(Station Blackout)
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Station Procedure A-50, Procedure for Conduct of Station Training
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Preoperational Test Procedure 2P-2.2 125/250 V (Div. I, II) DC Safeguard Power System FSAR Table 14.2-4
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Standard Review Plan (SRP) Section 14.2 This item was reviewed in IR 50-353/88-31 and left open pending action on three areas.
Closure of those three issues has betn achieved as follows:
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NRC Safety Evaluation is provided in SER Section 14, as TMI I
item I.G.1 is specifically called out in the acceptance criteria in the SRP.
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. Verification of Reactor. Core Isolation Cooling;(RCIC) operr -
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ability with non-RCIC batteries disconnected is provided by'
preoperational test 2P-2.2, which verifies that Motor Con-r trol Center 20D201 (RCIC.DC power supply) is fed only from
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th+. Division I batterv.
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Station blackout training is' incorporated into requalifi-
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. cation' training for all Reactor Operator, Senior Reactor Operator and Shift Tech,1 cal Advisor personnel, as cbcu-mented fn Appendix B to station procedure A-50.
This item is closed.
d.
(Closed) I.A.1.2.
Shift Supervisor responsibilities.
Supple-ment 1 to the Limerick Safety Evaluation Report approved PECos means for achieving compliance based upon a review of drvt administrative procedure A-7, ' Shift Ope ations." This i
'..e was reviewed in NRC inspection report 50-353/88-21 and left open.
pending verificatica that all shift supervisors-had completed a Kepner-Tregoe Course in Decision Making, as committed to in PECo response to FSAR question 630.27.
The final four personnel com-pleted this training in March and April 1989, as documented.in Limerick Plant Divi. ion Memo OPS-0807, Barnshaw to Firth, dated April 18, 1989. This item is closed.
e.
-(Closed) I.C.5.
Feedteck of operating experience Supplement 1 to the Limerick SER ac.
ed the Limerick program based upon the
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tracking system.identt ' ed in draft administrative procedure NGD-A-5, Procedure for Review and Utilization of Operating.Ex-7erience Information.
In June 1988, a revised Operating Ex-perience Assessment Program (0EAP) was promulgated in NGS-0XX.Y, 10terim Nuclear Group Administrative Procedure for Operating
'Exoerience Assessment Program. This revised program became ef-fective July 1, 1988.
Disposition, tracking and closure of OEAP items is parformed in accordance with AG-40, Control of LGS Operating Experience Assessment Program Items and Commitments.
This item is closed.
f.
(Closed) I.C.1.
Short-term accident and procedure review. This item was previously reviewed in Inspection Report 50-353/88-25, and left open pnding generation of revised TRIP procedures which include Un;t 2 specific information.
The inspector re-viewed the minutes of Plant Operations Review Committee (PORC)
meeting 83-130.
This meeting of PORC approved revisions to.five
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TRIP procedures which were necessary to permit use of the pro-cedures on either unit.
This item is closed.
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4.0 Surveillance /Special Observations (61726, 64704) (Unit 1 Only) i
During this inspection period, the inspector reviewed in progress-surveil-lance testing as well as completed surveillance packages.
The inspector verified that surveillance were performed in accordance with licensee approved procedures and NRC regulations.
The inspector also verified that instruments used were within calibration tolerances and that quali-fied technicians performed the surveillance.
The following surveillance were reviewed:
4.1 Unit 1 ST-4-095-921-1 Division I 125/250 VDC Safeguards Battery 19 Month Inspection RT-1-078-420-0 Battery Room HVAC Test 4.1.1 The inspector witnessed portions of the battery ' charger heat run test for the 1A2 charger. The test is performed per surveillance test ST-4-095-921-1 and was being per-formed to retest newly installed control circuit cards and to accomplish the 18 month technical specification (TS)
surveillance requirement. TS 4.8.2.1.C.4 requires that at least once per 18 months the licensee verify that the charger can supply 300 amps at a minimum of 132 volts for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During the test performance the resistor load bank circuit was deenergized temporarily and then restored.
The test was resumed and a total of eight hours of operation at 300 amps was accomplished.
The inspector questioned the validity of the test results, since during the time when the load bank was deenergized the heat gene-ration was reduced which potentially reduces the peak tem-peratures to which the charger components would be subject if the test is run for a continuous eight hour period.
Also the inspector noticed the battery charger cabinet door was open during the test thus potentially providing addi-tional cooling beyond thht which would be present under normal operation.
Since the chargers contain electronic control circuitry it is important that the peak cabinet temperature be obtained during the test so that proper operation of these circuits can be verified under worst case conditions.
These concerns were discussed with licensee personnel who agreed that it would be difficult to quantify the effects of a tc porary loss of load (i.e. non-continuous eight hour run) or the effect of opening the door during the test.
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3g The licensee repeated the t st with the cabinet doors closed and a continuous ei
,t hour run.
The results were satisfactory.
The test p cedures will also be revised to clarify the conditions tr ae maintained to obtain valid results.
5.0 Maintenance Observations (62703) (Unit 1 Only)
5.1 Unit 1 The inspector reviewed the following safety related maintenance ac-
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tivities to verify that repairs were made in accordance with approved
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procedures, and in compliance with NRC regulations and recognized
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codes and standards. The inspector also verified that the riplace-ment parts and quality control utilized on the repairs were in com-pliance with the licensee's QA program.
890277 Replacement of IBCA2 Battery Charger Control Board
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Replacement (Interim Modification 095-001)
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880431 Division III 18 Month Battery 095-002 Inspection / Maintenance
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RHRSW Pump Discharge Check Valve Repairs NA D-13 Emergency Diesel Generator Governor
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Troubleshooting
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8881773 Main Steam Safety Relief Valve PSV-041-1F013H 8902189 Replacement / Repair
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890?l87 Main Steam Safety Relief Valve PSV 041-1F013L 8881776 Replacement / Repair 5.1.1 Main Steam Safety Relief Valve Work On. March 31, the inspector was contacted by a contractor
employee who had a concern that two nf the main steam safety relief valves (MSRVs) had been damaged during in-stallation and were being utilized without the damage being repaired. The MSRVs involved were PSV-041-1F013H and PSV-041-1F013L.
The inspectors reviewed the maintenance procedures and work documentation associated with these jot and also inter-viewed members of the licensee mainter ance and quality con-trol staff.
The damage consisted of scoring on the main valve body which was caused when the pilot assemblies were being re-moved and rotated to achieve the proper configuration for solenoid valve mounting.
The pilot cartridge on the PSV-041-1F013L valve was also bent during the work. The damage
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was caused during the disassembly of the pilot cartridge l
assemblies by improperly rigging them from the valve main L
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If the pilot cartridge assembly is not lifted ver-l-
tically binding occurs in the base bore area of the valve.
Maintenance procedure M-041-007, Maintenance Procedure for Replacement of a Main Steam Relief Valve Pilot Stage As-sembly, includes specific requirements (as procedure " cat.-
l tions") on how to lif t the pilot stage assembly to prevent damtge and that the pilot stage assembly must be handled carefully to avoid any damage to the three inch portior which inserts into the base bore. Contrary to the proce-
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dure a hydraulic jack was also used to disassemble the l
PSV-041-1F013L valve which exerted excessive force on the
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components causing the bending of the p1?ot cartridge and resulted in scoring the bore.
When the PECo maintenance technical staff identified the valve problems, the valves were examined and dispositioned as fol!ows:
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The pilot cartridge assembly for PSV-041-1F013L was
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replaced with a Unit 2 assembly.
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Work instructions were issued to rcmove the burrs from
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the PSV-041-1F013L valve base bore and to verify the as left bore dimension is 3.000 to 3.005 inches.
The snap ring was straightened to remove burrs and it j
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was reinstalled on the PSV-041-1F013L pilot assembly, j
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The PSV-041-1F013H pilot assembly was measured and
imperfections were removed such that measurements were
' within vendor tolerances.
Snap rings and snap ring i
grooves were cleaned and reinstalled.
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In addition to the associated MRFs the inspector also re-viewed the following:
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M-041-005, Maintenance Procedure for Replacement of Main Steam Relief Valves
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M-041-007, Maintenance Procedure for Replacement of a Main Steam Relief Valve Pilot Stage Assembly Various Quality Control Inspection Reports (QCIRs)
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generated during the valve work Maintenance Work Instructions related to the
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PSV-041-1F013H and -L valve work
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Administrative Procedure A-26, Procedure for Plant Maintenance using the Maintenance Request Form Procedure NQA-24, Control of Hardware Nonconformances
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(Converted to Nuclear Group Administration Procedure NA-03-N001)
l The failure to follow procedure M-041-007 is a violation t
of TS 6.8.la which requires that written procedures be es-
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tablished, implemented and maintained covering the activi-ties referenced in Appendix A of Regulatory Guide 1.33,
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Revision 2, February 1978.
Section 9 of Appendix A to Regulatory Guide 1.33 requires procedures for the perform-ance of maintenance on safety-related equipment-i (50-352/89-09-01).
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The inspector also reviewed the licensee administrative l
procedures for documenting and resolving equipment non-conformances.
Station procedures direct personnel who dis-cover equipment problems to initiate an equipment trouble-tag (ETT) or a corrective maintenance request for (CMRF).
In this job a CMRF was already in use and the additional actions taken to resolve the scoring and bent pilot assem-blies were documented as changes to the CMRF. Although the
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valve deficiencies were corrected the root cause for the valve damage was not addressed.
The QCIRs written-during the work did not document the problem and under the current procedures the QC inspector cannot generate a nonconform-ance report using the present nonconformance system.
In order to ensure resolution of the valve deficiencies the QC inspector withheld signoff of the cleanliness verification hold point in M-041-007 which delayed the job.
The NRC inspectors expressed a concern that th-noncon-formance control system is difficult to understa i and in this case failed to identify a significant deficiency which required further root cause analysis and coriective ac-tions.
It was also noted that the Limerick Quality Assur-ance (QA) plan and procedure NA-03-N001 are not consistent in that the-QA plan states that the Plant Operations Review Committee (PORC) or the Independent Safety Engineering Group (ISEG) will review deficiencies to determine if they are significant whereas NA-03-N001 procedure states that the NQA department will accomplish this review.
The lic-ensee QA department is reviewing the nonconformance proce-dure and is planning to revise it to make it easier for the plant staff to utilize.
)
This allegation was not substantiated as the valves were
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properly repaired. However, the licensee's approach to i
maintenance of these valves and corrective action for damage caused during maintenance, as documented above, identified a violation of procedures and a quality control concern.
6.0 Review of Periodic and Special Reports (90713)
Upon receipt, the inspector reviewed periodic and special reports.
The review included the following:
inclur. ion of information required by the NRC; test results and/or supporting information consistent with design
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predictions and performance specifications; planned corrective action for-resolution of problems, and deportability and validity of report informa-tion. The following periodic report was reviewed:
Unit 1 Monthly Operating Report - March 1989
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The inspector had no questions regarding this report.
7.0 1.icensee Event Report Followup (90712, 92700) (Unit 1 Only)
The inspector reviewed the following LERs to determine that deportability requirements were fulfilled, that immediate corrective action was taken, and that corrective action to prevent recurrence was accomplished in ac-cordance with technical specifications.
7.1 Unit 1 88-031, Revision 2, continuing Review of Fire Protection Evaluation Report. This revision identifies the reason certain instrumentation (reactor vessel water level) could be lost during a fire and the licensee's actions to prevent recurrence. This event was discussed in report 50-352/89-03. The inspector reviewed the cause and cor-rective actions which are summarized as follows.
Root Cause:
1.
Lack of detailed procedures used in performing the safe shutdown analysis.
2.
A misunderstanding and misapplication of the detailed regulatory requirements.
Corrective Actions:
The corrective actions included an extensive series of ongoing studies. These studies have been discussed with Region I management l
and the results of the studies will be inspected when completed. The inspector has no further questions at this time.
,89-002, Revision 1;89-008, Revision 1 and 89-009 Revision 1 and 2 also contain updates on cable separation findings which will be ad-dressed when the aforementioned studies are completed.89-016, Manual Engineered Safeguards Feature (ESF) Actuation of Re-fueling Secondary Containment Isolation nd Standby Gas Treatment System due to Dropped Fuel Pin.
This LER describes the manual actu-ation of the refueling secondary containment isolation system and the start of the standby gas treatment when a fuel pin was dropped during fuel reconstitution activities.
The fuel pin was retrieved, found undamaged and returned to its proper location.
The ESF actuation was
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i a precautionary measure instituted in the event that the pin did:rup-
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ture.
The systems were returned to normal and fuel reconstitution
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I continued after discussions with the fuel handling crew. A procedure change was made to address p,oper actions to.be taken in the event of a dropped fuel pin. Additionally, the fuel reconstitution procedure was reviewed and found to be adequate.
Refueling technicians were
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counseled in its use in order to prevent an additional accidental
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disengagement of a pin from the grapple when the tie rod was un-i screwed from the bottom tie plate of the fuel bundle. The recon-j-stitution of the fuel was completed without further incident. The J
inspector has no further questions at this time.
>
89-017, Drywell Hydrogen Mixing System may be Unavailable to Mitigate i
the Consequence of an Accident. This event was discussed in inspec-
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tion report 50-352/89-03 and the inspector has no further comments at l
this time.
i 89-018, Inadvertent Start of 'C' Emergency Service Water (ESW) Pump.
This LER describes starting "C" ESW pump during the installation of
a modification that would provide local starting capability of the
"C" pump.
The modification did not properly describe the scope of the work and the blocking coordinator failed to rack out and tag the supply breaker to "C" ESW pump.
During the performance cf wire
. changes the pump started, which alerted the operator, who stopped the pump. The pump was lined up to its respective system and no damage occurred.
Construction personnel had discussions with the operators and were redirected by the operators to restore the wiring to the pre-modification configuration when a second pump start occurred.
Again the operator stopped the pump.
The proper blocking was sub-sequently established, and the modification was completed and tested properly.
The construction personnel involved were counseled in the use of administrative procedure A-14, " Procedure for Control of Plant Modifications." The inspector had no further questions regarding this event.89-019.
Certain electrical equipment required for operation was found to be lacking environmental qualification. This was discussed in inspection report 50-352/89-05 and was left open pending the lic-ensee's corrective actions. The licensee has performed modification 5977 which adds the environmental seals and low point drains to the safety related equipment which includes temperature elements, limit switches, solenoid valves and pressure switches. The inspector has no further questions at this time.89-020. Missed one hour fire watch inspections of certain fire pro-tected areas. This LER describes missed fire watch inspections of rooms 542 (Auxiliary Equipment Room) and 540 (Remote Shutdown Panel Room).
The cause of the event has been attributed to the inoperable card readers and the response of the security guard with the proper key in excess of one hour by 17 minutes. At the time, the unit was
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in a refueling mode and the fire detectors were not~ required.
The licensee has revised security procedures to ensure the availability of the proper keys. The inspector had no further questions regarding this event.
I 89-02.
Refueling floor isolation due to low differential pressure
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March 29, 1989.
This event is discussed in section 2.1.2 of this report. The inspector had no further comments.
8.0 Review of Licensee General Employee Training (GET)
The inspector attended and observed the GET requalification training at Limerick on March 30, 1989.
Training objectives were written out in de-tail and clearly presented.
Training on hazardous chemical materitis and heat stress was given.
Numerous mishaps or near misses in the industry were discussed to emphasize a point.
The inspector also noted that the Nuclear Plant Rules had been recently revised to incorporate the new re-quirement of not announcing the presence of an NRC inspector coming on site. No concerns were identified with the requalification training.
9.0 Technical Specification Review (Unit 2 Only) (371301}
Proof and Review Technical Specifications for Unit 2 were issued March 22, 1989. This version was compared page-by page against Unit 1 TS and re-viewed for consistency with the as-built plant and to verify that approved i
surveillance test procedures correctly implement the test frequencies and setpoints.
This review was a continuation of that started in inspection report 50-353/89-11. This review concentrated on the Reactor Protection System (RPS) and the two issues left unresolved in inspection report 50-353/89-11. The unresolved issues are documented in section 3 of this i
report. The documents reviewed are listed in Attachment 3.
Instrument setpoints, test frequencies and instrument response times for the RPS were found to be properly implemented in appropriate surveillance test proce-dures.
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10.0 Assurance of Quality, During this inspection period the inspectors questioned the assurance of quality regarding the disposition of the damage to the Main Steam Safety Relief Valves. Damage had occurred to the valves during maintenance and the licensee made prompt and proper repairs to the valves.
However, the root cause of how the damage had happened was not addressed.
The inspec-tors assessed the nonconformance control system and found it was difficult to understand and in this case failed to assure identification of the sig-nificant deficiency.
The inspectors had discussions with senior members of the QA department and as a result the department is planning to sim-plify the nonconformance system for easier utilization by the plant staff.
For more information regarding this event see section 5.1.1 of this re-port.
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3g 11.0 Exit Interview The NRC resident inspectors discussed the issues in this report throughout the inspection period, and summarized the findings at an exit meeting held with the Plant Manager, Limerick Generating Station, on April 28, 1989.
No written inspection material was provided to licensee representatives-during the inspection period.
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ATTACHMENT 2
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DOCUMENTS REVIEWED Letter D. Grace (BWR Owners Group) to A. Thadani (NRC) "NRC Bulletin No. 88-07 Supplement 1", dated January 26, 1989.
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Lesson Plan lqR-88-05-F " CORE: BWR Power Oscillation (LaSalle Event)", dated 12/21/88.
Simulator Exercise Guide (SEG-0013) " Recirculation Pump Trip from High Power and Subsequent Recovery", dated 7/19/88.
Letter J. Gallagher (PECO) to NRC " Limerick Generating Station and Peach Bottom Atomic Power Station Response to NRC Bulletin 88-07 Supplement 1", dated March 7, 1989.
Letter J. Gallagher (PECO) to C. Rossi (NRC) "NRC Bulletin No. 88-07 Peach Bottom Atomic Power Station and Limerick Generating Station", dated 9/15/88.
Memorandum P. Duca to W. Alden " Response to NRC Bulletin No. 88-07, Supplement 1", dated 1/25/89.
Memorandum P. Duca to W. Alden " Response to NRC Bulletin No. 88-07", dated 7/15/89 GP-2 " Normal Plant Startup", dated 3/7/89.
GP-3 " Normal Plant Shutdown", dated 1/26/89.
GP-4 " Rapid Plant Shutdown to Hot Shutdown", dated 1/26/89.
ON-113 " Loss of RECW", dated 1/26/89.
ON-114 " Loss of Stator Water Cooling Runback", dated 2/25/88.
OT-100 " Reactor Low Level", dated 7/15/88.
OT-102 " Reactor High Pressure", dated 1/23/89 OT-103 " Main Steamline High Radiation", dated 1/27/89.
OT-104 " Unexplained Reactivity Insertion", dated 1/26/89.
OT-112 " Recirculation Pump Trip", dated 2/14/89.
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ST-6-107-880-1 "APRM and LPRM Noise Level Determination", dated 7/14/88.
RE-201 " Reactor Maneuvering Plan Approval", dated 11/14/88.
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ATTACHMENT 3 Reactor Protection System-Technical Specification Review Table 2.2.1-1, Bases 2.1.3, 3/4.1.4, Figure 3.2.1-1, Figure 3.2.1-2, Figure 3.2.1-3, Figure 3.2.1-4, Figure 3.2.1-5, Table 3.3.1-1, Table 3.3.1-2, Table 4.3.1.1-1, Table 1.1, Table 1.2, Bases Figure B 3/4.3-1 Final Safety Analysis Report Table 6.3-4, Table 7.2-2, Table 15.0-2 Procedures Reviewed ST-2-001-402-2:
RPS and EOC-RPT-Turbine Stop Valve-Closure Calibration Test (ZS-01-204A,ZS-01-2048,ZS-01-204C,ZS-01-204D)
ST-2-001-421-2:
RPS and EOC-RPT-Turbine Control Valve Fast Closure Trip System 011 Pressure-Low, Division IA, Channel Al Calibration / Functional (PS-01-202A) (Control Valve 3)-
ST-2-001-422-2:
RPS and. E0C-RPT-Turbine Control Valve Fast Closure Trip System 011 Pressure-Low, Division IIA, Channel A2 Calibration / Functional (PS-01-2028) (Control Valve 4)
ST-2-001-423-2:
RPS and EOC-RPT-Turbine Control Valve Fast Closure Trip System 011 Pressure-Low, Division IB, Channel B1 Calibration / Functional (PS-01-202C) (Control Valve 1)
ST-2-001-424-2:
RPS and EOC-RPT-Turbine Control Valve Fast Closure Trip System 011 Pressure-Low, Division IIB, Channel B2 Calibration / Functional (PS-01-2020) (Control Valve 2)
ST-2-041-416-2:' RPS-Main Steam Isolation Valve-Closure, Division IA, Channel A1 -Calibration / Functional Test (ZS-41-222A, ZS-41-2228, ZS-41-228A, ZS-41-228B)
ST-2-041-417-2:
RPS-Main Steam Isolation Valve-Closure, Division IB, Channel B1 Calibration / Functional Test (ZS-41-222A, ZS-41-222C, ZS-41-228A, ZS-41-228C)
ST-2-041-418-2:
RPS-Main Steam Isolation Valve-Closure, Division IIA, Channel A2 Calibration / Functional Test (ZS-41-222C, ZS-41-222D, ZS-41-228C, ZS-41-2280)
ST-2-041-419-2:
RPS-Main Steam Isolation Valve-Closure, Division IIB, Channel B2 Calibration / Functional Test (ZS-41-222B, 25-41-2220, ZS-41-2288, ZS-41-228D)
ST-2-041-420-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IA, Channel A1/A Calibration / Functional Test (RE-41-2N006A, RISH-41-2K603A)
ST-2-041-421-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IB, Channel B1/B Calibration / Functional Test (RE-41-2N0068, RISH-41-2K603B)
ST-2-041-422-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IIA, Channel A2/C Calibration / Functional Test (RE-41-2N006C, RISH-41-2K603C)
ST-2-041-423-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IIB, Channel B2/D Calibration / Functional Test (RE-41-2N006D, RISH-41-2K6030)
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l ST-2-041-616-2: RPS-Main Steam Isolation Valve-Closure, Division IA, Channel l
Al Functional Test (ZS-41-222A, ZS-41-228A, ZS-41-2228, ZS-41-2288)
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ST-2-041-617-2:
RPS-Main Steam Isolation Valve-Closure, Division IB, Channel B1 Functional Test (ZS-41-222A, ZS-41-228A, ZS-41-222C, ZS-41-228C)
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ST-2-041-618-2:
RPS-Main Steam Isolation Valve-Closure, Division IIA, Channel
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A2 Functional Test (ZS-41-222C, ZS-41-228C, ZS-41-2220, ZS-41-228D)
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ST-2-041-619-2:
RPS-Main Steam Isolation Valve-Closure, Division IIB, Channel i
B2 Functional Test (ZS-41-222B, ZS-41-228B, ZS-41-222D, ZS-41-228D)
ST-2-041-620-2: RPS and NSSSS-Main Steam Line Radiation-High, Division IA, Channel A1/A Functional Test (RISH-41-2K603A)
ST-2-041-621-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IB, Channel B1/B Functional Test (RISH-41-2K603B)
ST-2-041-622-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IIA, j
Channel A2/C Functional Test (RISH-41-2K603C)
ST-2-041-623-2:
RPS and NSSSS-Main Steam Line Radiation-High, Division IIB, Channel B2/D Functional Test (RISH-41-2K6030)
ST-2-042-445-2:
RPS and NSSSS-Reactor Steam Dome Pressure-High, Division IA, Channel A Calibration / Functional Test (PT-42-2N078A, PIS-42-2N678A, PS-42-2N679A)
ST-2-042-446-2:
RPS and NSSSS-Reactor Steam Dome Pressure-High, Division IB, Channel B Calibration / Functional Test (PT-42-2N0788, PIS-42-2N678B, PS-42-2N679B)
ST-2-042-447-2:
RPS and NSSSS-Reactor Steam Dome Pressure-High, Division IIA, Channel C Calibration / Functional Test (PT-42-2N078C, PIS-42-2N678C, PS-42-2N679C)
ST-2-042-448-2:
RPS and NSSSS-Reactor Vessel Water Level-Low, Level 3, Division IA, Channel A Calibration / Functional Test (LT-42-2N080A, LIS-42-2N680A)
i ST-2-042-454-2:
RPS and NSSSS-Drywell Pressure-High, Division IB, Channel B Calibration / Functional Test (PT-42-2N0508, PIS-42-2N6508)
ST-2-042-647-2:
RPS and NSSSS-Reactor Steam Dome Pressure-High, Division IIA, Channel C Functional Test (PIS-42-2N678C, PS-42-2N679C)
ST-2-042-652-2: RPS and NSSSS-Reactor Vessel Water Level-Low, Level 3,
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Division IIB, Channel D Functional Test (LIS-42-2N680D)
ST-2-042-656-2:
RPS and NSSSS-Drywell Pressure-High, Division IIB, Channel D Functional Test (PIS-42-2N6500)
ST-2-047-409-2:
RPS-Scram Discharge Volume Water Level-High, Division IIA, Channel C Calibration / Functional Test (LT-47-2N012C, LISH-47-2N601C)
ST-2-047-603-2:
RPS-Scram Discharge Volume Water Level-High, Division IIB, Channel D Calibration / Functional Test (LSH-47-2N013D)
ST-2-047-407-2:
RPS-Scram Discharge Volume Water Level-High, Division IA, Channel A Calibration / Functional Test (LT-47-2N012A, LISH-47-2N610A)
ST-2-074-404-2:
IRM A Calibration Test i
ST-2-074-413-2: APRM B Calibration / Functional Test ST-2-074-608-2:
IRM Channel A Functional Test ST-3-074-505-2: TIP Calibration of LPRMs ST-6-001-660-2: Main Turbine Stop Valve RPS & EOC-RPT Channel Functional Test ST-6-001-765-2: Main Control Valve Exercise & RPS Channel Functional Test ST-6-001-306-2:
Channel Al and A2 RPS Manual Scram Channels Functional Test ST-6-097-300-2:
Rx Mode Switch Functional Test
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i ST-6-107-884-2:
Neutron Monitoring System Overlap Verification on Shutdown ST-6-107-886-2:
Neutron Monitoring System Overlap Verification on Shutdown ST-6-071-100-2:
Reactor Protection System Logic System Functional / Simulated Automatic Actuation ST-2-001-801-2:
RPS and NSSSS Turbine Stop Valve-Closure, Division IA, Channel Al Response Time Test (ZS-01-204A) (Main Stop Valve #3)
ST-2-001-806-2:
RPS and EOC-RPT-Turbine Control Valve Fast Closure Trip System 011 Pressure-Low, Division IIA, Channel A2, Response Time Test, (PS-01-2028) (Control Valve 4)
ST-2-041-800-2:
RPS-Main Steam Line Isolation Valve-Closure, Division IA, Channel Al Response Time Test (HV-41-2F022A, B; HV-41-2F028A, B)
ST-2-042-829-2:
RPS cnd NSSSS-Reactor Steam Dome Pressure-High, Division IB, Channel B Transmitter Response Time Test (PT-42-2N078B)
ST-2-042-834-2:
RPS and NSSSS-Reactor Vessel Water Level-Low, Level 3, Division IIA, Channel C Tranmitter Response Time Test (LT-42-2N080C)
ST-2-042-931-2:
RPS and NSSSS-Reactor Steam Dome Pressure-High, Division IIB, Channel D Response Time Test (PIS-42-2N678D)
ST-2-042-932-2:
RPS and NSSSS-Reactor Vessel Water Level-Low, Level 3, Division IA, Channel A Response Time Test (LIS-42-2N680A)
ST-2-074-802-2: APRM C Response Time Test
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