IR 05000293/1988034
| ML20235S238 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/14/1989 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235S236 | List: |
| References | |
| 50-293-88-34, GL-88-14, NUDOCS 8903070032 | |
| Download: ML20235S238 (17) | |
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U.S. NUCLEAR REGULATORY. COMMISSION.
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REGION'I Docket No.:
50-293 h
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Report No.:
50-293/88-34 Licensee:
Boston' Edison Company
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800:Boylston Street'
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Boston, Massachusetts 02199
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Facility:
Pilgrim Nuclear Po'wer Station s
Location:
Plymouth, Massachusetts-Dates:
November 14.- December 26, 1988 Inspectors:
C. Warren, Seriior Resident Inspector T. Kim, Resident Inspector C. Carpenter,. Resident Inspector R. Barkley, Reactor Engineer M. Kohl, Reactor Engineer P. Drysdale, Reactor Engineer G. Bryan, Nr.:. Contractor G.-Bethke, NRC Contractor
Approved by:
D/y-8d A. Randy Bloug W hief.
Date Reactor Projects Section No. 3B Division of Reactor Projects
_I_nspection Summary:
Areas Inspected:
Routine resident inspection of plant operations,. radiation.
i protection, physical security, plant ever.ts, maintenance, surveillance,. Part 21 reports and outage activit 45.
Principal licensee management contacted are
- listed in Attachment I to this report.
Results:
Strengths:
The' licensee's licensee event reports (LER) are clear, concise and provide a thorough analysis of the event, causes, corrective actions and safety implica-tions (Section 5.0).
Weaknesses: None Observations:
Throughout the inspection period, the licensee demonstrated a thorough and safety conscious approach to followup of events and issues.
8903070032 890217 PDR ADOCK 05000293
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ITABLE OF CONTENTS Page 1.
-S umma ry o f f a c i l i ty Act i v i t i e s.............................-
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Routine Periodic -Inspections'(Modules: 61726',' 62703,,
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71707,'71709, 71881).....................
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3.
-Review of Plant Events (Modules 37700,1 62703, 62726,
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71707,.71709)..........................................
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a.
ESF Actuation-Due:to ' Low Reactor Vessel WaterLLe' vel b.
Radioactively Contaminated Liquid Spill c.
Potentially Inoperable.ControliRoom High Efficiency
' Air Filtration System d.
Secondary Containment. Leak Rate Test' Failure e.
Senior Licensed Operator Leaving Control Room-Without. Proper Relief 4.
10 CFR Part-21; Report by.Limitorque Concerning Melamine.
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-Torque' Switches (Module 92703)..........................
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5.
Review of Licensee Event Reports (LERs)'-(Module 90712)....
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' Management Meetings (Module:30703)........................
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.AttachmentLI'- Persons Contacted
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j DETAILS 1.0 Summary of Facility Activities
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The plant has been shut down for maintenance and program improvements j
since April 12, 1986.
The reactor core was completely defueled on
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February 13, 1987, to facilitate extensive maintenance and modification of plant equipment. The licensee completed fuel reload on October 14, 1987.
Reinsta11ation of the reactor vessel internal components and the. vessel head was followed by completion of the reactor vessel hydrostatic test.
During this report period, the licensee performed routine maintenance and surveillance tests'and continued to prepare the plant for restart.
Effective December 1, 1988, Mr. Edwin J. Wagner assumed the position of Director of Nuclear Engineering.
He replaced Mr. J. Howard, Vice Presi-dent, Nuclear Engineering, who retired on December 1, 1988.
Effective December 26, 19.88, Mr. Leon Olivier assumed the duties of Chief Operating Engineer. The former Chief Operating Engineer requested and received re-assignment to a Nuclear Watch Engineer position.
NRC inspection activities during this report period included a:
1) rou-tine health physics inspection of the licensee's radiological controls program during the week of November 28,1988 (Inspection Report 50-293/
88-35); and 2) observation of the licensee's annual onsite emergency pre-pardness exercise on December 13,1988 (Inspection Report 50-293/88-36).
On December 9,1988, the NRC Commissioners met in Rockville, Maryland to hear from officials of local communities near the Pilgrim plant with emergency planning responsibilities and the Commonwealth of Massachusetts on the status of emergency planning in the area around the plant.
The Commission also heard from the NRC staff concerning the stctus of the Pilgrim facility.
On December 21, 1988, the NRC Commissioners votea unanimously to endorse the NRC staff's proposal to permit the licensee to restart the Pilgrim Nuclear Power Station provided the staff is satisfied that the licensee is ready to proceed with the Power Ascension rogram.
o The licensee's NRC-accepted Power Ascension Program included several NRC holdpoints; i.e.,
points at which the licensee has agreed to obtain NRC Regional Administrator approval before proceeding.
2.0 Routine Periodic Inspections The inspectors routinely toured the facility during normal and backshift hours to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security and radiological control measures.
Inspections were conducted between 10:00 p.m. and 6:00 a.m. on December 2 and 8, 1988, for a total of five hours and on the weekends and holidays of
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November 19, 20, 26' an'd 27, December 17,18' and 26,1988,..for a total of
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' 49.5. hours. Ongoing work activities were monitored to verify that they were being conducted in accordance with approved administrative.and tech-nical. procedures and that proper communications with the control room staff had been established. The inspector coserved valve, instrument and electrical equipment lineups in the field to ensure that they were con-sistent with system operability requirements and operating-procedures.
During tours of the control room the inspectors verified proper staffing, access ' control ana operator attentiveness.
Adherence to procedures and limiting conditions for operations (LCO) was evaluated.
The inspectors examined equipment lineup and operability, and status of' control room annunciators.
Various control room logs and licensee documentation were reviewed.
The inspector observed and reviewed outage, maintenance and p'roblem.inves-tigation activities to verify compliance - with regulations, procedures, codes and standards.
Involvement of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest requirements and deportability were' assessed.
The inspector observed surveillance and post-work tests to verify perform-ance in accordance with approved procedures and LCO's, collection of valid test, results, removal and restoration of equipment and deficiency review and resolution.
Radiological controls were observed on a routine basis during the report-ing period.
Conformance to standard industry radiological work practices, radiological control procedures, and 10 CFR Part 20 requirements v&s observed.
Independent surveys of radiological boundaries and random sur-veys of nonradiological points throughout the facility were taken by the inspector.
Checks were made to determine whether security conditions met regulatory requirements, the physical security plan and approved procedures.
Those checks incl uded security staffing, protected and vital area barriers,
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personnel identification, access control, badging and compensatory i
measures when required.
No discrepancies were identified by the inspectors.
3.0 Review of Plant Events The inspectors followed up on events occurring during the period to deter-mine if licensee response was thorough and effective. Independent reviews.
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l of the events were conducted to verify the accuracy and completeness of
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licensee information.
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Engineered Safety Feature (ESF,) Actbation Due to Low Reactor Vessel a.
Water Leve_i_
On December 3,'1988, at 3:04 a.m.,
the licensee received a full reactor scram, Groups II, 'III, VI ' and secondary containment isola-tior,s and actuation of the standby gas treatment system on low reactor vessel water level (less than +9 inches). The reactor scram and isolations occurred while the licensee was placing the "A" loop of the residual heat removal (RHR) system in the shutdown cooling mode. The licensee notified the NRC via ENS at 4:15 a.m. in accord-ance with 10 CFR 50.72(b)(2)(ii).
The event involved. a sequence of actions that resulted in incomplete filling of the RHR system,. before unisolating it from the reactor.
Following completion of local leak rate testing (LLRT) and the per-formance of hydrodynamic (valve seat leakage) testing on RHR shut-down cooling valves MD-1001-47 and 50, the shutdown cooling suction piping had been left in a partially drained configuration. The shut-
down cooling discharge piping had been left filled, through the open
M0-1001-50 valve up to valve MO-1001-47, which was closed.
Conse-quently, valve M0-1001-47 separated the filled side of the - system (discharge piping) from the partially drained side of the system (suction piping).
On December 3, after completion of the LLRT the licensee cleared the tags from the shutdown cooling system and began procedure 2.2.86, Residual Heat Removal, to place the "A" loop of the RHR system in the shutdown cooling mode.
The sequence of steps in the procedure in preparing for the shutdown cooling (SDC) mode consists of flushing (7.2.3); filling and venting the SDC suctior, piping (7.2.4.1); fill-ing and venting the SDC discharge piping (7.2.4.2); opening the SDC suction line isolation valve (MO-1001-50) (7.2.9) and opening the SDC suction line isolation valve (M0-1001-47) (7.2.10).
In an effort to expedite the work, the licensee decided to perform flushing of the suction piping in parallel with filling and venting of the. discharge piping. Upon completion of flushing of the suction
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piping, the auxiliary operators involved under the direction of the Nuclear Operations Supervisor, began an on-shift break.
When the licensee returned to the procedure, they recommenced at the point in the procedure following section 7.2.4.2 (fill and vent of discharge pipe).
The fill and vent of a portion of the suction piping not included in the previously perforwed flush had been bypassed because of the concurrent flushing and filling operations.
When valve M0-1001-47 was opened, water from the reactor vessel flowed into the partially drained shutdown cooling suction piping, allowing reactor vessel water level to drop below the low reactor vessel water level setpoint (+9"), causing the ESF actuations.
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Two contributing causes. of - this _' event we're11dentifiea; by' the: licen '
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First, the procedure was ' not? followed in the sequence Las writ-~
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should have been filled"and vented, followed byzthe fill and venting of. the. discharge piping. In attempting to perform:two paragraphs' of:
the procedure concurrently, the paragraph that;would have filled andl
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vented the partially drained _- suction line was missed. ' Second,1there c existed a. general misunderstanding. by' some licensee personnel that g
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procedures did not need to be followed sequentially. Contributing to:
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this-misunderstanding.was the lack of guidance..in licensee procedure <
1.3.34,~ _ Conduct of Operations, as to ' how procedures' are.i to be-
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Licensee corrective actions include the' following:
1).0perations'-
-procedure 1.3.34 is being revised to strengthen-the guidance provided
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for. the, performance-of_ procedural operations (e.g.; in. the identified,
. sequence), 2)1the RHR sy' stem procedures are being revised to provide a caution that will identify the portions of the. procedure that have-
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to be completed prior to opening SDC isolation valves M0-1001-50 or M0-1001-47, and 3) a night r order was issued as an interim measure to p'rovide guidance to ' operators relative to the performance of proced -
ural operations in the identified sequence.
At thet close of the inspection'. period, the RHR' system procedures and. Operations procedure 1.3.34 were. in draft form.
The -inspector considers the ongoing
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corrective actions to be appropriate and had no further questions.
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Radioactively Contaminated Liquid Spill
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On November 16,;1988, the licensee discovered that radioactively con-I-
taminated' liquid waste.had;been spilled inside a process building and.
- that some.of the liquid had-leaked outside the building. The licen -
see estimated that about 2300 gallons were spilled with about 100 gallons exiting the _ process building.
Prompt licensee response pre-
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' vented the contaminated water from escaping the' site's owner con-trolled area.
A-detailed review of the event was. conducted by a-region-based radiation safety inspector and the results are docu-mented in inspection report 50-293/88-35.
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- Potentially In0perable Control Room High Efficiency Air Filtration L
. System On November 38, 1988, the licensee.. reported to the NRC.via the~ ENS,
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' that the control room high efficiency. air filtration system (CRHEAF)
may be inopercble under a degraded voltage condition.. The:11censee's analysis indicated that the minimun: required output of-3.43 KW ' for the first' bank of heaters in.each CRHEAF train 1may not be achieved at-
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the minimum anticipated bus voltage of 420 VAC.
The CRHEAF has two trains and each train has four banks of heaters.-
The first bank. of. heaters activates on either ~1ow' inlet.-temperature-or high humidity. The other three banks of heaters are designed for maintaining the comfort level in the control-room. ; The M censee.
implemented a temporary modi'fication on November 18, 1988, to connect the T1 phase of the second heater to the T1 phase of the first heater
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of..both trains to increase the capacity of the first bank. heaters.
The licensee's analysis indicated that an output of 4.79 KW at 480
.VAC and 3.92 KW. at 420 VAC would be achieved ~ for the first bank. >
heaters with the_new configuration. The post modification test per--
formed on. November 18, 1988, confirmed the above analysis. The Tech-nical Specification requires a total -of 14 KW for all four banks of -
heaters. The licensee's previous surveillance records indicated that there.has been no problem in meeting this requirement, The licensee's engineering department is in the process of.' reviewing the ' design capacity of. the heaters with the vendor.
The ' licensee also plans to perform another test with installed in-line ammeters-to obtain more accurate output readings. The inspector will follovup
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on the licensee's actions in a future inspection under existing WRC
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outstanding item 88-31-01, d.
Secondary Containment Leak Rate Test Failure
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The licensee. performed 'a secondary containment leak rate test with a revised test procedure on December 22, 1988 and determined that the test results failed to meet the station technical specification.
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required acceptance criteria of 0.25 inches of water vacuum at 4000 l
cubic feet per minute (CFM) air flow through the ' standby gas treat-
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ment system. The test procedure had been recently revised to perform the test without inflating the seal on the secondary containment
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trucklock inner door.
The licensee had determined that the air supply to the seal was not seismically qualified during their follow-up to Generic Letter 88-14, " Instrument Air Supply System Problems Affecting Safety-Related Equipment." With the seal deflated, there was approximately a 3/4 inch gap between the seal and the trucklock
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The test data indicated 0.18 inches of water vacuum at 3700 l
CFM air flow through the standby gas treatment system. The licensee made an ENS notification to the NRC at 11:25 p.m. on December 22 to
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notify the NRC of the test failure. The previous leak rate test per-formed on September 21, 1988, with the seal inflated had indicated acceptable results.
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As a corrective action, the licensee installed flexible rubber-flaps around the trucklock door, in. addition to the existing seal, as a permanent fixture. Another secondary containment leak rate test was scheduled to be performed on December 28, 1988.
The licensee's review of the station instrument air system is ongoing in response
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The inspector will review licensee actions in this area in a future inspection report.
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Senior Licensed Operator Leaving Control Room Without Proper Relief On Friday, December 2,1988, at approximately 3:00 p.m., the on-duty Watch Engineer, an NRC licensed senior reactor operator (SRO), left the site without being relieved of his licensed duties. The Pilgrim station minimum license requirements for the number of SRO's required onshift is two and at the time the Watch Engineer left he was ona of four SR0's present in the control room. One of the two extra SR0's present established himself as the duty Watch Engineer end assumed'
those responsibilities for the remainder of the shift.
The licensee immediately suspended the Watch Engineer's station 2 cess.
Extensive medical staff evaluation concluded that the indi-vidual was fully fit for duty but overreacted to external stress factors.
The interviews conducted with the individual established that he was aware of his responsibilities as a licensed operator and that he felt those responsibilities were met because there were numerous SR0 licensed individuals in the control room at the time of his departure.
The licensee plans to take disciplinary action against the individual and to assign him to non-licensed duties. A formal program to evalu-ate the individual's performance will be conducted. If the licensee later decides to return the individual to licensed duties, the basis for that conclusion will be discussed with NRC prior to the indi-vidual assuming those duties.
Although the issues that instigated this event appeared to be spec-ific in nature and related only to the Watch Engineer in question, licensee management decided to conduct an in-depth review of factors that could affect the working conditions of on-shift personnel. This independent review was performed by three individuals who subse-quently compared observations and perceptions. All on-shift super-visory personnel were interviewed and no generic issues were identi-fied that would have negatively impacted operational performanc _ -
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The inspectors as well as NRC. Region I management, followe'd' closely _
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the licensee's efforts throughout this event followup and have found i
them to be. prompt.and thorough. The licensee has taken firm disci-p11 nary action and has repeatedly stressed the importance of operator
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conduct to all members of the operations staff.
The inspector con-siders that the licensee-adequately. addressed all areas of concern and has no further questions.
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. 4.0: 10 CFR Part 21 Report by Limitorque Concerning Melamine Torque Switches ~
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. On November-3,1988, Limitorque Corporation notified lthe licensee of the issuance of a 10 CFR Part 21 notification concerning: failures of Melamine torque switches installed in Limitorque motor operators supplied to: the'
Washington Public Power Supply System (WPPSS). Limitorque determined that these failures, specifically cam binding, represented a common mode fail-ure resulting from post mold shrinkage of Melamine and could potentially affect the valve actuator's ability to perform its safety function.
The.
- 1 motor operator failures experienced at WPPSS were caused by the binding of-
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the cam on -the torque switch shaft. The binding caused the cam to open the torque switch contacts at the lowest switch setting regardless of the actual torque switch setting.
Melamine torque switches were primarily used in valve actuators qualified for harsh environments.
Limitorque identified that SMB-000 actuators with serial numbers less than 354839 and -SMB-00 actuators with serial numbers lower than. 233218 could have been supplied with Melamine torque switches. The SMB-000 and SMB-00 Melamine torque switches are easily identified by_ a white or gray-colored insulating material.
The licensee reviewed recent inspection records on safety-related Limit-orque motor operated valves (MOVs) and also performed a field inspection to determine if the torque switches installed in the motor operators have Melamine material; two MOVs in the RHR system and a MOV in the service water system were found with Melamine torque switches. As recommended by Limitorque, the licensee replaced these torque switches with environmen-
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tally qualified (EQ) fiberite torque switches. The inspector reviewed the maintenance work plan and the procedure 3.M.3-24.2, Limitorque Type SMB-000/00 Motor Operator Overhaul, for the torque switch replacement work and determined that they were technically adequate and thorough. Motor Oper-ated Valve Acceptance Testing (MOVATs) testing was performed on these valves which verified that the torque switches operated properly.
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5.0 Review of Licensee Event Reports (LER's)
LER's submitted to NRC:RI were reviewed to verify that the details were clearly reported, including causal description and adequacy of corrective
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action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated and whether
the event warranted onsite followup.
The following LER's were reviewed:
LER No./
Subject Event Date 88-01-00/
Automatic actuations of portions of the primary and 01/06/88 secondary containment isolation and standby gas treat-
ment systems due to a failed relay coil in the actua-tion logic. The cause was due to a random failure of the coil in logic relay 16A-K57, creating a current fault which blew the logic system power fuses.
The failed coil was a type CR120A relay manufactured by General Electric.
The inspector's review of this actuation is documented in Inspection Report 50-293/
87-57.
88-02-00/
Full scram trip signal during a surveillance test, 01/17/88 resulting in incomplete automatic actuations.
The cause of the actuation was procedural inadequacy in that the procedure did not contain sufficient instruc-tions or cautions.
The inspector's review of this actuation is described in Inspection Reports 50-293/
87-57 and 50-293/88-12.
Although the LER identifies the cause for the Reactor Protection Systtm (RPS) trip signal to be procedural inadequacy in that the pro-cedure did not contain sufficient instructions or cautions to alert the technician, the inspector's review of this event in Inspection Report 50-293/87-57 identified that the level instruments (LI-263-59 A&B)
were incorrectly installed (e.g., the sensing lines were reversed). No mention of this cause was made in the original LER. Followup review by the inspector in
Inspection Report 50-293/88-12 indicated that the reactor vessel level instruments were incorrectly i
installed due to an error in the configuration draw-l ings which were issued as part of Field Revision
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Notice No. 62 to Plant Design Change (PDC) 85-07. As of the end of this report period, a supplement to this l
LER had not been issued as committed to in the origi-nal LER. The inspector will review the LER supplement during a subsequent inspection to determine if all root causes of this event are identified by the licen-
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'9 LER No./
Subject Event Date
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88-03-00/
Low setpoints of degraded grid voltage relays due to 01/30/88 errors in the computer model used for analysis. The cause of the setpoint error was due to an incorrect assumption made in the computer model used for the degraded grid voltage analysis.
The inspector's review of licensee actions is described in Inspection Reports 50-293/88-07, 50-293/88-23 and special elec-trical team Inspection Report 50-293/88-08.
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88-04-00/
Unplanned actuation of portions of the primary and 02/02/88 secondary containment isolation systen, and auto start of the standby gas treatment system. The cause of the actuation was inadequate instructions for work being performed in a Primary Containment Isolation Syster/
Reactor Building Isolation System (PCIS/RBIS) logic panel.
A secondary cause for the actuations, not identified in the LER but in the inspector's review of
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the event, was the failure to perform adequate tech-I nical review of the PDC.
This failure resulted in violation 88-07-01.
The inspector's review of the actuation is contained in Inspection Report 50-293/
88-07.
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88-05-00/
Automatic actuation of portions of the primary con-02/02/88 tainment and secondary containment isolation systems i
and auto start of the standby gas treatment system.
The cause of the actuations was the failure of the coil in a logic relay that was part of the inboard PCIS/RBIS logic circuitry.
The inspector's review of this actuation is described in Inspection Report 50-293/88-07.
88-06-00/
Anticipated Transient Without Scram (ATWS) Division II 02/03/88 trip signal and subsequent scram signal.
The root cause of the ATWS trip signal had not been identified at the time of submittal of the LER but was being investigated.
A supplemental report was not yet issued at the close of this report period.
The inspector's initial followup of the trip signal is described in Inspection Report 50-293/88-07.
88-07-00/
Unexpected actuation of portions of the secondary 02/23/88 containment and standby gas treatment systems.
The cause of the actuation was the removal of a fuse from the power supply of a logic circuit during planned preventive maintenance.
Root cause was lack of ade-quate communication and the use of an inadequate pro-cedure 'speci fied for work preparation.
The inspec-tor's review of this event is discussed in Inspection Report 50-293/88-07.
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LER-No./.-
Subject
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Automatic. closing: of a primary containment system:
02/26/88 Group 6 isolation ' valve.
The automatic closing Tof
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valve M0-1201-5 occurred whenithe breaker supplying 125_ VDC power to the : valve operator wa_s closed. The i
root cause of the actuation was-a. momentary flow fluc =
tuation.in the Reactor Water Cleanup System. Immedi-ate ~ inspector followup is documented in Inspection Report-50-293/88-07, 88-09-00/
tMissed surveillance during December 1983 -for
"B" s
03/04/88.
emergency diesel generator. = The exact cause of the -
missed surveillance could not be determined.
Inspec-tor review of this event is' documented in Inspection -
Report 50-293/88-07.
88-10-00/
Reactor. water cleanup system spurious ' isolation. The 03/11/88 cause of the actuation was a blown fuse.in the inboard PCIS logic circuitry which occurred during the removal.
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of - a temperature switch.
Root cause. was personnel e rro r, The inspector's review of this actuation' is contained in Inspection Report 50-293/88-12.
188-11-00 Inadvertent actuation.' of the reactor bui_1 ding' isola-03/31/88 tion ' control system.
The causetof the ' actuation was-attributed to ' personnel error in 1that ther operator incorrectly performed a portion of a routine surveil-lance of the reactor building' refuel floor. radiation.
monitors, resulting in. the failure to properly. reset the downscale trip for - two of the channel s.
The inspector's review of this event is discussed in Inspection ~ Report 50-293/88-12.
88-12-00/-
Automatic start of a Reactor Building Closed Cooling 04/25/88 Water (RBCCW) system. pump.
The cause of the pump start was a'ccelerated ' wear induced failure of the coupling in the RBCCW pump. The inspector's review of this event is documented in Inspection Report 50-293/
88-19.
88-13-00/
Inadvertent manual start of the "B" emergency diesel 04/25/88 generater.
The cause was attributed to personnel error in that the manua'l start switch was mistakenly pushed instead of an annunciator reset switch.
Review of this event is discussed in Inspection Report 50-293/88-19.
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LER No./.
Subject
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Event Date 88-14-00/
Reactor water. cleanup isolation.
Root cause was the 04/26/88'
failure of the coil in a logic relay, causing the cir-cuit to deenergize.. The inspector's review 'is con-tained in Inspection Report 50-293/88-19.
88-15-00/
Inadvertent actuation of portions of the primary con-05/04/38 tainment, secondary containment and standby gas treat-ment systems. The cause of the actuation was attrib-uted to personnel error in that the wrong fuse was removed _ from its circuit during planned work.
The inspector's immediate followup is contained in Inspec-tion Report 50-293/88-19.
88-16-00/
Actuation of portions of the secondary containment and 05/17/88 standby gas treatment systems. The cause of the actu-ation was the removal of a fuse from a logic circuit during relay coil replacement due to personnel error.
The inspecttr's review of this event is documented in Inspection Report 50-293/88-19.
88-17-00/
Crack in yoke portion of a residual heat removal sys-06/08/88 tem valve. At the time-of the' LER, the cause of the crack and indications in the yoke portion of the valves had not yet been determined positively, but an updated report would be submitted after completion of the investigation.
The licensee subsequently deter-mined the cause for the cracked yoke on the one valve and indications on the other valve yoke to be from two i
causes:
(1) deficiencies in the design and (2) fabri-i cation of the valve yoke and excessive motor operator thrust.
Detailed inspector review of the valve yoke deficiencies is contained in Inspection Reports 50-293/88-25, 50-293/88-27 and 50-293/88-31.
i 88-18-00/
Full reactor protection system scram signal during 07/05/88 instrument calibration.
The cause of the full scram signal was a procedural weakness in that the procedure changes were not consistent with the Intermediate Range Monitor's (IRM) vendor manual and inadequate
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administrative controls.
Inspector followup of this event is contained in Inspection Report 50-293/88-25.
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Subject Event Date 88-19-00/
Full scram si.gnal resulting from ATWS testing.
A 07/08/88 partial scram signal had been inserted as part of the planned test of the ATWS Division II circuitry.
A scram was caused due to unforseen, but normal, leakage into the scram discharge volume which resulted-from the ATWS signal.
An ' inadequate review of the test procedure had failed to identify that a full scram signal would occur as a consequence of the test. In-spector review of this event is contained in Inspec-tion Report 50-293/88-25.
88-20-00/
Automatic start of
"B" emergency diesel generator 07/11/88 (EDG).
The cause for this event was personnel error in that the Nuclear Operating Supervisor failed to recognize that the
"B" EDG emergency run/stop switch was not included in the tagout sheet prior to work to be performed.
Inspector. followup of this event is documented in Inspection Report 50-293/88-25.
88-21-00/
Deficient solenoid valves installed in safety-related 07/19/88 applications.
The cause was an incorrect assumption regarding a failure of the air system in that the condition of. air overpressurization was not con-sidered. A failure of the air system was assumed to only result in lower system pressure; this assumption would be incorrect if an air system pressure regulator failure resulted in higher pressure downstream of the regulator.
The inspector's review of this item was documented in Inspection Report 50-293/88-27.
88-22-00/
Full scram signal due to licensed operator error.
The 10/10/88 cause of the full scram signal was operator error in that the operator opened the circuit breaker (SA-CB1A)
that provided power to the Reactor Protection System (RPS) chaanel
"A" logic circuitry in order to reset the breaker. The inspector's review of this event is documented in Inspection Report 50-293/88-33.
88-23-00/
Full isolation of secondary containment and auto start
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10/12/88 of the standby gas treatment sy stem.
The cause of
this event was operator error in that the "A" refuel i
floor radiation monitor was not reset after restoring the normal power supply to the "A" RPS buses. Proced-I ure 2.2.79,
" Reactor Protection System", was also determined to be inadequate in the instructions pro-
vided to the operator.
This event was reviewed in Inspection Report 50-293/88-33.
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13-LER No./
Subject Event Date 88-24-00/-
Primary - Containment (Group I) Isolation. 'The isola-10/17/88'
tion occurred due to high water -level in the reactor vessel that was caused by a mechanical feedback cam for. a feedwater valve. that became disconnected and therefore gave the control room a false " closed" indi-cation. This event was reviewed by' the inspectors in Inspection Report 50-293/88-33.
88-25-00/
Full reactor scram, primary containment end secondary 12/08/88 containment isolation and - actuation of standby gas:
treatment system. The cause of the actuation was low reactor vessel water level as a result of operator error. The procedure to be used to place the residual heat removal system in operation was not followed verbatim.
The inspector's review of this event is detailed in section 3.0 of this Inspection Report.
The inspector's review of licensee LER's determined that the LER's ade-quately complied with 'the requirements of 10 CFR 50.73 in that they were timely, the description of the event in each case was clear, concise and the LER accurately described the sequence of events. With some exceptions described above, root cause analyses generally appeared to be thorough in that contributing factors to the event were also detailed and corrective actions appeared adequate to prevent recurrence. In each case the inspec-tor noted that the LER addressed the safety implications of the event on the health and safety of the public and included a review for similar previous events.
In December, 1988 the licensee identified 22 LER's requiring a supplemen-tal report (update to the original LER) be submitted to the NRC. Fourteen of these are for the time. period between 1986 and 1988; the remaining eight are for the time period between 1980 and 1985.
Discussion with the licensee indicated that, although the LER's have been tracked on their LER tracking system, the LER's had not been updated in a timely manner due to the lack of priority that had been assigned to update and submit the supplemental reports.
The licensee has subsequently developed a schedule for submittal of those supplements.
The licensee expects to issue supplemental reports to the i
14 LER's initiated between 1986 and 1988 prior to March 31, 1989. For the i
LER's initiated between 1980 and 1985 requiring supplemental reports, only
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one report has been assigned a date for submittal.
Discussions with the I
licensee indicated that priority was given to issue supplements to those LER's written after the plant shutdown in early 1986.
The inspector I
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considers' this.to-be acceptable.. However, the inspector questioned the licensee's decision not t'o > assign the remaining-LER's (1980 to 1985) which.
require updatesa an expected submission date. ~ The Linspector considered-that a commitmentto complete these LER updates would ' ensure their timely submittal.
The licensee subsequently established' an ' expected submission
'date of September 30..1989, for the - remaining LER supplemental reports.
The" licensee' has' added any LER's l requiring supplements to their computer tracking system which tracks NRC commitments. Tracking system reports are issued to management 'on a weekly basis.
The inspector will followup.
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licensee timeliness on issuance of-LER supplemental reports in a subse-quent inspection report.
6.0 Management Meetings At periodic intervals during the course of the inspection period, meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident' inspectors. A final inspection exit interview was conducted on January -27,1989. - No written material was given to the licensee that was not previously available to the public.
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l Attachment I to Inspection Report 50-293/88-34
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-Persons Contacted R. Bird, Senior Vice President - Nuclear ~
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Highfill, Station Director R. Anderson, Plant Manager E. Kraft, Plant Support Department Manager A. Morisi, Acting-Outage and Planning Department Manager D,.-Swanson, Nuclear Engineering Department-Manager-J. Alexander, Plant Operations.Section Manager
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J. Jens,. Radiological Section Manager J. Seery, Technical Section Manager R.: Sherry, Maintenance Section Manager P. Mastrangelo, Chief Operating Engineer
.D. Long, Security.Section Manager W. Clancy Systems. Engineer Division Manager F. Wozniak, Fire: Protection Division Manager S. Dasgupta, Control' Systems Division Manager R. Kirven, Power System Division Manager
- Senior licensee representative present at the exit meeting.
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