IR 05000483/1987032
| ML20234F340 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/31/1987 |
| From: | Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20234F320 | List: |
| References | |
| 50-483-87-32, IEB-85-003, IEB-85-3, IEIN-87-004, IEIN-87-008, IEIN-87-010, IEIN-87-012, IEIN-87-10, IEIN-87-12, IEIN-87-4, IEIN-87-8, NUDOCS 8801110432 | |
| Download: ML20234F340 (17) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-483/87032(DRP)
Docket No. 50-483 License No. NPF-30 Licensee:
Union Electric Company Post Office Box 149 - Mail Code 400 St. Louis, MO 63166 Facility Name:
Callaway Plant, Unit 1 Inspection At:
Callaway Site, Steedman, Missouri Inspection Conducted:
October 4 through December 4,1987 Inspectors:
B. H. Little C. H. Brown s
Y Approved By m J. M. Hin s, N
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} Reactor Projects Section 1A Date Inspection Summary Inspection from October 4 through December 4, 1987 (Report No. 50-483/87032(DRP))
Areas Inspected:
A routine, unannounced safety inspection of Licensee Event Reports (LERs); plant operations; Engineered Safety Feature (ESF) systems; radiological controls; maintenance; surveillance; fire protection; emergency preparedness; security; quality programs; outages; regional requests; IE bulletin followup; and training.
_Res ul ts :
Of the 14 areas inspected, no violations were identified in 13 areas.
Two violations were identified in the remaining area (failure to meet the required response time for a safety injection signal - Paragraph 2.b (1);
failure to perform required stroke testing of valves EF-HV-0087 and 0088 -
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Paragraph 2.b(2)).
However, in accordance with 10 CFR 2, Appendix C, Section V.A., Notices of Violation were not issued.
The two violations were of minor safety significance.
I sa 8901110432 871231 ADOCK050g3 PDR G
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DETAILS 1.
Persons Contacted D. F. Schnell, Vice President, Nuclear G. L. Randolph, General Manager, Callaway Plant
- J. D. Blosser, Manager, Callaway Plant l
C. D. Naslund, Manager, Operations Support
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A. P. Neuhalfen, Manager, Quality Assurance J. R. Peevy, Assistant Manage", Technical Services W. R. Campbell, Assistant Manager, Nuclear Engineering
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M. E. Taylor, Superintendent, Operatior,
D. E. Young, Superintendent, Maintenan;e
- W. R. Robinson, Assistant Manager, Operations and Maintenance R. R. Roselius, Superintendent, Health Physics T. P. Sharkey, Supervisor, Ccmpliance G. J. Czeschin, Superintendent, Planning and Scheduling W. H. Sheppard, Superintendent, Outages G. R. Pendegraff, Superintendent, Security i
- L. H. Kanuckel, Supervisor, QAPS Engineer R. D. Affolter, Superintendent, Systems Engineering G. A. Hughes, Supervisor, Independent Safety Engineer Group J. V. Laux, Superintendent, Technical Support, Quality Assurance J. C. Gearhart, Superintendent, Operations Support, Quality Assurance I
- J. J. Cassmeyer, Quality Assurance Engineer l
- Denotes those present at one or more exit interviews.
In addition, a number of equipment operators, reactor operators, senior reactor operators, and other members of the quality assurance, quality control, operations, maintenance, health physics and engineering staffs were contacted.
2.
Inspection of Licensee Event Reports (92700)
Through direct observations, discussions with licensee personnel, and review of records, the following licensee event reports were reviewed to determine that deportability requirements were fulfilled, that immediate corrective action was accomplished, and that corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications (T/Ss).
The LERs listed below are considered closed.
a.
LERs 87-007-00 and 87-026-00:
Containment purge and/or control room ventilation isolations resulting from personnel errors.
The events appear to be isolated occurrences with regard to the nature of tasks and personnel involved.
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(1) LER 87-007-00:
On May 4, 1987, an ESF acuation (control room ventilation isolation [CRVI]) occurred when the test probes of a digital voltmeter (DVM) were incorrectly placed across terminals in a solid state protection system (SSPS) cabinet.
The test probes were incorrectly placed during troubleshooting of a control problem with a containment isolation valve.
To check the containment isolation signal to this valve, a utility Instrumentation and Controls (I&C) technician was to perform a continuity check across terminals 3 and 4 on an SSPS relay.
However, he placed the test probes across terminals 13 and 14 on the relay.
When he tested for continuity, terminals 13 and 14 shorted through the DVM and the CRVI occurred.
The required ESF equipment performed as designed.
l The inspector determined that this matter received appropriate licensee attention and action.
Action to prevent recurrence included shop meetings with the I&C technicians who covered the event and stressed precautions during troubleshooting.
Additionally, progressive discipline was administered to the I&C technician involved.
l LER-87-007-00 is considered closed.
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(2) LER 87-026-00:
On September 16, 1987, an ESF actuation (containment purge isolation-[ CPI] and CRVI) occurred during reactor vessel head venting following initial reactor coolar'
system (RCS) draindown in preparation for refueling activi+1 2s.
Two subsequent CPI actuations occurred during restoration from
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the initial CPI.
l Following the third event, the licensee performed a ventilation-system line-up check, which found the containment shutdown damper closed.
This condition caused an insufficient dilution flow to and corresponding high radiation levels at radiation monitors GT-RE-22 and GT-RE-33.
The licensee's review determined that a non-licensed utility equipment operator improperly placed WPA tag #22 on PG20GA1 Circuit #22 instead of PG20GA1 Circuit #38.
Based on a discussion with the person involved, the misplacement was a mental transposition error.
This improper placement resulted in the failure of the damper at the containment shutdown purge exhaust fan outlet in the closed position. When the shutdown purge was initiated, this damper i
remained closed, thus preventing exhaust airflow.
Without airflow through the shutdown purge exhaust, the gas vented from the reactor vessel concentrated in the exhaust duct, and caused the trip setpoints for GT-RE-22 and GT-RE-33 to be exceeded.
Consequently, the CPI and CRVI ensued.
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The inspector's reviews of this matter found that appropriate controls were in place for venting the reactor vessel, which included task planning and task instructions containing. health physics (HP) recommendations.
Venting of the vessel was directed to the purge exhaust by the use of tubing, which was immediately closed upon receipt of the ESF actuation signals, thus precluding increased exposures to personnel in the containment.
As the event resulted in containment isolation, the event posed no threat to the public.
The licensee took appropriate corrective action, which included supervisory counsel of the utility equipment operator involved in this event.
This is considered an isolated occurrence.
LER 87-026-00 is considered closed.
No violations or deviations were identified.
b.
LERs 87-006-00 and 87-027-00:
Violations of Technical Specifications.
The events were assessed for recurrence and pregram weaknesses.
The events appear to be isolated occurrences.
No generic weaknesses were noted.
(1) LER 87-006-00:
Safety injection (SI) signal response time in excess of T/S limits due to erroneous vendor data.
On April 13, 1987, the licensee was notified by Westinghouse Electric Corporation of a potential inaccuracy in the T/S relating to the response times associated with an SI signal.
A review of Table 3.3-5 of the Callaway Plant Technical Specifications determined that the va've realignment times had not been included.
Normal charging to the RCS is accomplished with the suction of the charging pumps aligned to the volume control tank (VCT).
Upon receipt of an SI signal, the pump suction is realigned to the refueling water storage tank (RWST).
To assure that the water supply to the charging pumps is not lost during this realignment, an interlock prevents the VCT valves from closing until the RWST valves are fully open.
Since the VCT is at a higher pressure than the RWST, the source of water to the charging pumps will be the VCT until the VCT valves are closed.
For one of tne SI initiators (steam line pressure - low), the accident ana'ysis assumes that borated water is injected into the RCS witlin a specified time, and therefore the time l
required for this valve realignment should have been included in the respanse time table of the Technical Specifications.
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T/S Limiting Condition for Operation (LCO) 3.3.2 requires that I
the Engineered Safety Features Actuation System (ESFAS)
instrumentation channels and interlocks shown in Table 3.3-3 shall be operable with their trip setpoints set consistent with the values shown in the trip setpoint column of Table 3.3-4 and with response times as shown in Table 3.3-5.
A test was performed to determine if the response times in the Technical Specifications could be met if the valve realignment times were included.
The most limiting SI response time is associated with the steam line pressure - low SI.
The required system response is 24 seconds including the Diesel Generator
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(D/G) start time, or 12 seconds excluding the D/G start time.
The test determined that the realignment occurred in 14.45 seconds, thus causing the response time to exceed the value
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specified in T/S Table 3.3-5.
Based on a safety evaluation performed b," Westinghouse, the licensee determined that there was sufficient margin in the safety analysis to increase the response times to include the
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i valve realignment time.
On April 16, 1987, the licensee initiated an emergency T/S change to revise T/S Table 3.3-5 to include tne valve realignment time.
On May 4, 1987, the NRC approved and issued the change as Amendment No. 22 to the Operating License.
This matter is considered to be an isolated event resulting I
from the licensee's failure to include the valve realignment time in the Technical Specifications.
The event was of minor safety significance, as the event was bounded by previous Final Safety Analysis Report (FSAR) analyses and posed no
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The licensee's failure to meet the specified response-time prior to the issuance of Amendment No. 22 to the Operating License is a violation of T/S 3.3.2.
This violation meets the tests of 10 CFR 2, Appendix C, Section V. A. ; consequently, no Notice of Violation will be issued, and LER-87-006-00 is considered closed.
(2) LER 87-027-00:
Essential service water (ESW) valves not surveilled as required by T/S 4.0.5.
On September 16, 1987, the licensee's Quality Assurance (QA)
department identified that two valves in the ESW system had not been surveilled as required by T/S 4.0.5.
QA documented the missed surveillance (valve stroke tests) in Incident Report (IR) No.87-183.
The missed valve surveillance were identified as a result of the licensee's corrective action (surveillance program evaluation effort) in response to surveillance-related T/S violations and surveillance deficiencies previously identified by the licensee and documented in NRC Inspection Report No. 483/87005(DRP).
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Valves EF-HV-0087 and EF-HV-0088 (ESW Train A and Train B, respectively, to radiation monitor isolation valves) are included in the licensee's in-service test (IST) program.
The IST program specifies that the valves be stroke tested quarterly (every three months) and following maintenance prior to returning the valves to service.
These valves are normally open valves, designed to close automatically on the initiation of an SI signal to isolate non-safety related components from the ESW system.
Valves EF-HV-0087 and 0088 were removed from service (tagged shut) to support corrective maintenance on associated non-safety piping.
It appears that failure to perform the required surveillance resulted from errors in the implementation of controls relating to the removal and restoration of safety-related components.
T/S 4.0.5 surveillance requirements specify that in-service testing of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Boiler and Pressure Code.
Failure to perform the specified surveillance constituted a failure to meet the operability requirements of T/S LC0 3.7.4.
Verification of the operability of valve EF-HV-0087 required that the valve be stroke tested on or before May 2, 1986.
The valve was not tested until July 10, 1986.
Verification of the operability of valve EF-HV-0088 required that the valve be tested on or before April 21, 1986.
The valve was not tested until June 3, 1986.
T/S LC0 3.7.4 action requirements were not satisfied in either case. When stroke tested, both valves performed satisfactorily.
The corrective action taken by the licensee subsequent to this event but prior to identification of this event appears
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adequate to prevent recurrence.
Corrective action includes:
Worker's Protection Assurance (WPA) sheets are required to i
reference existing EOSL sheets.
Surveillance procedures may test multiple components.
As
an enhancement to the surveillance administrative program, new surveillance task sheets are issued to ensure the testing of components that were exceptions to the original surveillance, e.g.,
single component failure or not testable due to plant conditions.
The task sheet serves as an additional reminder to utility planning personnel or to other responsible departments.
A summary of this event has been provided to the licensed
utility shift supervisors, operating supervisors, and the shift technica' advisors for their review.
A design change has been developed to correct the clogging
of the ESW supply lines to the radiation monitors.
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The valves performed satisfactorily when tested, and the event posed n'o significant threat to the public or plant safety.
The surveillance deficiencies were discovered by the licensee's QA j
department as part of the licensee's efforts to improve the
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surveillance program.
The failure to perform the specified stroke testing of valves EF-HV-0087 and 88 in accordance with the licensee's IST program j
is a violation of T/S 4.0.5.a.
This violation meets the test of 10 CFR 2, Appendix C, Section V.A., consequently, no Notice of Violation will be issued.
LER 87-027 is considered closed.
3.
Plant Operations (71707)
a.
Operational Safety Verification Inspections were routinely performed to ensure that the licensee conducts activities at the facility safely and in conformance with regulatory requirements.
The inspections focused on the implementation and overall effectiveness of the licensee's control l
of operating activities, and the performance of licensed and l
nonlicensed operators and shift technical advisors.
The following i
items were considered during these inspections:
Adequacy of plant staffing and supervision.
- Control room professionalism, including procedure adherence,
operator attentiveness, and response to alarms, events, and off-normal conditions.
Operability of selected safety-related systems, including
attendant alarms, instrumentation, and controls.
Maintenance of quality records and reports.
- The inspections included direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and LCOs, and reviews of facility procedures, records, and reports.
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i The licensee's operating crews demonstrated their ability to I
effectively supervise and control refueling outage and plant restoration activities.
The inspectors observed pre-startup, j
initial criticality, and post-startup activities from the control room.
Initial criticality for fuel cycle 3 was achieved on November 6, 1987.
Prior to the completion of core physics testing, the plant was returned to cold shutdown on two occasions to perform corrective maintenance on a main steam isclation valve (MSIV) and on a power-operated relief valve.
Prior to the shutdown for repair of the MSIV, the MSIV was cycled.
During this activity a reactor trip occurred as the result of steam generator low water level.
i Subsequent plant cooldowns and heatups were performed without incident.
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The inspectors reviewed / observed performance of the following activities by operating crews:
Procedure No.
Activity 0TG-ZZ-00005 Plant Cooldown Hot Standby to Cold Shutdown OTG-ZZ-00007 Refueling Preparations, Performance and Recovery ODP-ZZ-00014 Mode Change Requirements ETP-ZZ-ST002 Initial Criticality E0P-ZZ-00007 Refueling Startup Testing b.
Off-shift Inspection of Control Rooms The inspectors performed routine inspections of the control room during off-shift and weekend periods; these included inspections between the hours of 10:00 p.m. and 5:00 a.m.
The inspections were conducted to assess overall crew performance and, specifically,
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control room operator attentiveness during night shifts.
The i
inspectors also reviewed the licensee's administrative controls regarding " Conduct of Operations" and interviewed the licensee's security personnel, shift supervisors and operators to determine if shift personnel were notified of the inspectors' arrivals on site during off-shif ts.
The inspectors determined that both licensed and non-licensed operators were attentive to their duties, and that the inspectors'
arrivals on site were unannounced.
The licensee has implemented appropriate administrative controls relating to the conduct of l
operations.
These include procedures which specify fitness for duty i
and operator attentiveness.
Personal radios and reading materials are prohibited.
No violations or deviations were identified.
4.
ESF System Walkdown (71710)
The operability of selected engineered safety features was confirmed by the inspectors during a walkdown of the accessible portions of several systems. The following items were included:
verification that procedures match the plant drawings, equipment conditions, housekeeping, instrumentation, valve and electrical breaker lineup status (per I
procedure checklist), and verification that locks, tags, jumpers, etc.
are properly attached and identifiable.
The following systems were walked down during this inspection period:
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"A" Emergency Diesel Generator
Component Cooling Water System
Essential Service Water System
Auxiliary Feed Water System
"B" Emergency Diesel Generator
Reactor Trip System
Main Steam Line Isolation
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125 Volt D.C. Power Source
AC Electrical Vital Power Source No violations or deviations were identified.
5.
Radiological Controls (71709)
The licensee's radiological controls and practices were routinely observed by the inspectors during plant tours and during the inspection of selected work activities.
The inspection included direct observations of HP activities relating to radiological surveys and monitoring, maintenance of radiological control signs and barriers, contamination, and radioactive waste controls.
The inspection also included a routine review of the licensee's radiological and water chemistry control records and reports.
The inspectors performed frequent plant tours during the refueling outage to assess the effectiveness of the licensee's HP program implementation relating to occupational radiation safety.
The inspectors observed that additional HP staffing was provided for the increase in work activities.
HP supervision attended work planning meetings and provided ALARA review and special training.
The licensee closely monitored personnel contamination incidents.
Radiological control barriers, signs and zones were maintained.
HP staff coverage was provided for special work activities and at hot particle buffer zones.
Good HP housekeeping practices were observed.
The inspectors observed that personnel entering, working in, and exiting radiological control areas generally displayed good radiological work practices.
The inspectors determined that the licensee had implemented effective radiological controls during the Cycle 2 refueling outage.
Inspection of the licensee's radiological controls and particles emphasized observation of the licensee's " hot particle" detection and prevention program.
This program included the development of a Hot Particle Action Plan, which was implemented to minimize the potential of personnel exposure to hot particles.
The plan was aimed at improved detection and prevention methods.
Actions taken to minimize the potential of hot particle exposures included special training, extra protective clothing, buffer zones established outside hot particle areas, decontamination, stay times, laundry monitors, additional personnel contamination monitors (PCM-1s), and an underwater vacuum system for the
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reactor cavity and spent fuel pools.
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The licensee's program for early detection of hot particles and prevention of personnel overexposure was effectively implemented.
There were no personnel overexposure during the refueling outage.
Health Physics records indicate the following:
A total of 72 hot particles were found this outage.
Twenty-nine of
the particles were fuel fragmants with 12 found on personnel.
The balance of the particles were corrosion products with nine found on personnel.
Twenty-one skin dose calculations were performed with 14 personnel
assigned skin exposure due to the hot particles / fuel fleas found on them.
No administrative limits were exceeded (7.5 rem allowed to the skin each quarter).
Calculated exposures ranged from 0 to 4.160 rem.
The average skin dose assigned due to fuel fragments was 2.083 rem; the highest exposure assigned was 4.160 rem.
No violations or deviations were identified.
6.
Maintenance (62703)
Selected portions of the plant maintenance activities on safety-related systems and components were observed or reviewed to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and standards, and the Technical Specifications.
The inspection included activities associated with preventive or corrective maintenance of electrical, I&C, and mechanical equipment and systems.
The following items were considered during these inspections:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved
procedures and were inspected as applicable; functional testing and/or-calibration was performed prior to returning the components or systems to i
service; parts and materials that were used were properly certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintained.
The reviewed maintenance activities included:
Work, Request No.
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WR-426535 Repack Main Steam Isolation Valve "B" WR-107163 Repack Pressurizer Power-operated Relief Valve Block Valve No violations or deviations were identified.
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7.
Surveillance (61726)
The inspectors reviewed or observed selected portions of the Technical Specifications required surveillance testing.during power operations.
Items which were considered during the inspection included whether adequate procedures were used to perform the testing, whether test instrumentation was calibrated, whether test results conformed with Technical Specifications and procedural requirements, and whether tests
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were performed within the required time limits.
The inspectors determined that the test results were reviewed by individuals other than the personnel involved with the performance of the tests, and that any deficiencies identified during the testing were reviewed and resolved by appropriate management personnel.
The inspectors' overview of the licensee's surveillance program showed that the licensee had developed and implemented a comprehensive and effective program.
The reviewed surveillance included:
Procedure No.
Activity 1SF-SE-00N36 Functional Test - Nuclear Instrument Intermediate Range-(N-36)
OSP-SB-C0001 Reactor Trip Breaker - P-4 Verification OSP-ST-01020 Shutdown Margin Calculations OSF-ST-13113 Personnel Air Lock Leak Rate Test OSP-ST-07084 Main Steam Isolation Valve Stroke Test ISL-SE-00N31 Loop-Nuclear:
Source Range N31 Instrumentation ISL-SQ-00Y64 Loop-Vibration:
Loose Parts Monitor ISP-SA-02414 Balance of Plant /E5FAS Trip Actuating Device Test ITL-BB-L1321 Reactor Vessel Level Train B OSP-ZZ-00001'
Control Room Shift and Daily Log Readings and Channel Checks MTE-ZZ-QA001 Motor Operated Valve Testing Using MOVATS No violations or deviations were identified.
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8.
Fire Protection (64704)
Implementation of the licensee's fire protection / prevention program was routinely assessed by the inspectors during plant tours.
The inspection included observation of housekeeping conditions, storage and control of combustible material, operability of fire protection / suppression systems, and fire brigade staffing and training.
One unannounced fire brigade drill was monitored during the inspection interval.
The response time of the brigade to assemble was satisfactory.
The manual fire equipment was noted to have been maintained; and at the completion of the drill, the equipment was returned. to readiness status.
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The housekeeping and control of combustible materials and flammable liquids were found to be satisfactory.
The review of selected surveillance for this area indicated that the surveillance were up to date.
The operability of the fire protection / suppression equipment and systems was maintained.
Fire watches and patrols were utilized in areas of the plant when required by the Technical Specifications.
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i No violations or deviations were identified.
9.
Emergency Preparedness (82301)
An inspection of emergency preparedness activities was performed to assess the licensee's implementation of the emergency plan and implementing procedures.
The inspection included monthly observation of emergency facilities and equipment, interviews with licensee staff, and a review of selected emergency implementing procedures.
No emergency preparedness drills were performed during this inspection interval.
No violations or deviations were identified.
10.
Security (71881)
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l The licensee's security activities were observed by the inspectors during j
routine facility tours and during the inspectors' site arrivals and i
departures.
Observations included the security personnel's performance
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associated with access control, security checks, and surveillance activities, and focused on the adequacy of security staffing, the security response (compensatory measures), and the security staff's attentiveness and thoroughness.
No violations or deviations were identified.
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11. Quality Programs and Administrative Controls Affecting Quality (35701)
An inspection of the licensee's quality programs was performed to assess the implementation and effectiveness of programs associated with management control, verification, and oversight activities.
The inspectors considered areas indicative of overall management involvement in quality matters, including the frequency of management plant tours and control room observations,'and management personnel's attendance at technical, planning / schedule, and committee meetings.. The inspectors attended onsite revir.w committee meetings and incident / event critiques and reviewed related documents,~ focusing on the licensee's root cause determinations and corrective actions.
The inspectors accompanied licensee management on monthly plant tours, which focused on quality activities and material conditions within the plant.
The inspection also
included a review of quality records and selected quality assurance audit and surveillance activities.
Performance in this area included the following major items:
a.
Callaway Plant Positive Moderator Temperature Coefficient On November 6, 1987, procedure ESP-ZZ-00009, Moderator Temperature Coefficient (MTC) Measurement at Zero Power, was performed to satisfy T/S surveillance requirement 4.1.1.3.
T/S 4.1.1.3 requires that prior to initial operation above 5% of rated thermal power after each fuel loading, the MTC shall be measured and determined to be less positive than 06K/K/ F for the all rods withdrawn,. beginning of
life cycle, hot zero thermal power condition.
The MTC was measure to be a positive 0.372 pcm/ F.
The plant was in Mode 2 (Startup) at 0% reactor power.
The predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods j
withdrawn condition has been calculated to be 6000 MWD /MTU.
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The inspector verified that the licensee submitted the special report to the NRC as required by T/S 3.1.1.3.a. Action a.3 and that the interim control rod withdrawal limits have been implemented, b.
Safety System Functional Assessment (SSFA)
The licensee has developed and implemented the SSFA program at Callaway.
The essential service water system will be the first safety system to undergo an actual assessment.
The program is scheduled to commence in January,1988.
The program provides assignment cf responsibilities and instructions for corrective
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action, reporting and evaluation of assessment fundings.
The assessment methodology will be similar to the NRC's Safety System Functional Inspection (SSFI) approach.
This includes an in-depth performance based assessment of a plant safety system rather than a program compliance approach.
The QA Division of the Quality Systems Department will maintain coordination and oversight responsibility for the initial SSFA.
Upon completion of the initial l
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SSFA, the results will be evaluated by Nuclear Function and Quality
Systems management for future application of this methodology.
j No violations or deviaticns were identified.
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Outages (71707, 61715)
The inspectors observed or reviewed licensee and contractor activities associated with plant outages.
The inspection focused on outage management program implementation, including planning, scheduling, and oversight activities.
The inspection included routine attendance of planning and scheduling meetings, direct observation of selected modifications, repair or testing of safety systems or components, and the review of quality records.
The management involvement was noted to be effective during the pre-outage planning and during the actual outage.
Management attended the shift turnover briefings conducted by the Shift Supervisors for all involved work groups.
The briefings maintained good communications between all work groups.
The ALARA program was one of the high priority iters in planning the work items for the outage.
During the recovery from the outage and preparations for plant restart, the 8-hour routine shift was reinstated.
The initial criticality of core 3 was monitored by the inspector.
The operating crews had received refresher training on the simulator for the initial startup.
During the startup of the reactor from the " Refuel 2" outage, selected portions of the following engineering tests were monitored:
ETP-ZZ-ST001 - Low Power Testing Data Acquisition
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ETP-ZZ-ST002 - Initial Criticality ETP-ZZ-ST004 - Boron Endpoint Measurement ESP-ZZ-00016 - Rod Drop Time Measurement ETP-ZZ-ST005 Rod Bank Worth Measurement Program procedure EDP-ZZ-00007, " Refueling Startup Test Program," was in use and was kept up to date.
The inspector made a tour of the reactor containment building prior to the final closecut of the containment.
All equipment and tools had been removed except for items that would be stored in special containers that
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are secured in place.
Radiological control barriers had been removed with the exception of the signs posting special areas within the containment building.
No unsecured items were noted, and the inspector considered containment housekeeping to be satisfactory.
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The inspector monitored selected local leak tests and valve function tests that verify containment integrity.
A portion of the routine sampling and identification of containment tendon wires was monitored.
The as-found containment leak rate was determined to be within the requirements.
The repair and retest calculations indicate that the containment leak rate was reduced to less than half of the allowable value.
The containment coolers and fans were noted to be operable.
The temperature control of containment, the penetration rooms, and other areas with temperature sensitive equipment was discussed in IR No. 483/87025 to document a regional request.
The inspector witnessed selected portions of the receipt and. inspection of new fuel and the placing of the fuel assemblies in the spent fuel storage pit racks.
The required personnel were present and the appropriate procedures and checklists were used.
The communications were considered adequate.
Only minor items were noted during the fuel assembly inspection.
Current levels of fission product activity indicate that the licensee's efforts to identity and replace leaking-fuel pins were effective.
Near the end of fuel cycle 2, the average dose equivalent I-131 (DEI-131)
was 4.83 E-2 pCi/gm.
This high DEI-131 was due to leaking fuel pins.
The RCS I-134 levels were also high, indicating that the fuel defects were open.
During Refuel 2, fuel assemblies were tested ultrasonically to determine leaking pins.
A total of 17 leaking fuel pins were found in 13 fuel assemblies.
Three of these assemblies were reconstituted by replacing the leaking pin in each assembly with a stainless steel pin.
The three reconstituted assemblies were reinstalled in the core.
After Refuel 2 the average DEI-131 was 3.97 E-3 pCi/gm.
No violations or deviations were identified.
13.
Reaional Requests (92701)
In a memo dated May 18, 1987, the inspectors were requested to perform a review of four selected IE Information Notices (IEINs).
a.
IEIN No. 87-04:
Diesel Generator fails test because of degraded fuel.
The review of this item showed no similar failures.
A fuel oil tank sample showed a high particulate concentration, and a temporary filter was installed and the tank recirculated for a short period of time before the 1987 spring outage.
The temporary filter and recirculation maintained the particulate concentration less than.the Technical Specification requirement of 10 mg/l total particulate; therefore, the diesel generator remained operable during this time.
During the outage the fuel oil was transferred to the auxiliary boiler fuel tank, and the diesel generator fuel oil tank and piping were cleaned and filled with new fuel oil.
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d
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The Technical Specifications require tnat the new fuel oil be
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sampled before it is added to the storage tanks and that the tanks be sampled monthly.
Tests and acceptance criteria are specified by ASTM-D-975-81 and D2276-78.
Thu licensee's program appears adequate to preclude the problems identified in the information notice.
b.
IEIN No. 87-08:
Degraded motor leads in Limitorque DC motor operators.
The review of this notice was reported in Inspection Report No. 50-483/87005 and closed.
c.
IEIN No. 87-10:
Potential for water hammer during restart of residual heat removal pumps.
This IN was applicable to BWR plants and sent to Callaway for information.
The site received the notice and reviewed it for applicability.
d.
IEIN No. 87-12:
Potential problems with metal-clad circuit breakers, General Electric Type AKF-2-25.
The Callaway site does not use this type of breaker in any functions similar to those noted in the information notice.
The site received the notice and reviewed it for applicability to the Callaway Plant.
No violations or deviations were identified.
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14.
IE Bulletin Followup (92703)
As requested by action item e. of Bulletin 85-03, " Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch i
Settings, " the licensee identified the selected safety related valves, i
the valves' maximum differential pressures and the licensee's program to assure valve operability in their letters dated May 14 and October 17, 1986, and March 5 and June 10, 1987.
The first three of these responses prompted requests for additional information which were contained in Region III letters dated August 1,1986, and February 13 and May 7,1987.
Review of Enclosure 2 of the June 10, 1987 response indicates that the licensee's selection of the applicable safety-related valves to be addressed and the valves' maximum differential pressures meet the requirements of the bulletin and that the program to assure valve operability requested by action item e of the bulletin is acceptable, subject to the clarifications contained in its August 18, 1987 letter.
The licensee's description of its post-maintenance program to ensure continued valve operability (action item d. of the bulletin) indicated a reliance on the ASME Boiler and Pressure Vessel Code Section XI testing.
Such a program meets the minimum requirements currently specified by the NRC for in-service testing of motor-operated valves and hence the minimum requirements of the bulletin.
However, the NRC Office of Nuclear Regulatory Research is currently reviewing the adequacy of relying solely on ASME Section XI testing to demonstrate valve operability.
There is a
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concern that stroke-time testing, which is done without differential pressure, does not demonstrate that the valve has sufficient margin to
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accommodate the additional loadings which will occur at the time of required valve operation.
Any changes prompted by this study could impact the licensee's post-maintenance program.
I The results of the inspections to verify proper implementation of this program and the review of the final response required by dction item f.
of the bulletin will be addressed in additional inspection reports.
i 15.
Training and Qualification Effectiveness (41400)
A performance evaluation of the licensee staff and contractors was performed to assess the effectiveness of the licensee's training and qualification program.
The following item:: were considered:
Program implementation (various work groups)
Understanding of work
Event and problem causal factors
Experience feedback
The inspection included personnel interviews, direct observation of training, operation, maintenance, and testing activities, observation of incident / event response, and a review of quality records and reports.
No violations or deviations were identified.
16.
Violations For Which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement.
However, because the NRC wants to encourage and support licensee initiatives for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.A.
These tests are:
(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and (5) it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violation.
Violations of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued are discussed in Paragraph 2.b.
17.
Exit Meeting (30703)
The inspectors met with licensee representatives (denoted under Persons Contacted) at intervals during the inspection period.
The inspectors summarized the scope and findings of the inspection.
the licensee representative acknowledged the findings as reported herein.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection.
The licensee did not identify any such documents / processes as proprietary.
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