IR 05000322/1988007
| ML20205Q264 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/20/1988 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205Q248 | List: |
| References | |
| 50-322-88-07, 50-322-88-7, NUDOCS 8811090175 | |
| Download: ML20205Q264 (11) | |
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W U b.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.:
50-322/88-07 Docket No.:
50-322 Licensee:
Long Island Lighting Company P. O. Box 618
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Shoreham Nuclear power Station Wading River, New York 11792 Inspection At: Wading River, New York
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Inspection Conducted: July 9, 1988 - September 15, 1988 Inspectors:
F. Crescenzo, Senior Resident Inspector R. Barkley, Reactor Engineer M. Kohl Reactor Engineer
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Approved By:
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l A. Randy 9f5 ugh, Chief Date
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Reactor Projects Section No. 3B
Division of Reactor Projects
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Inspection Summary:
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Arets Inspected:
Routine Resident Inspection of plant operations, radiation protection, security, plant events, maintenance, surveillance, outage activ-
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ities, and reports to the NRC.
Non-routine inspection activities were con-
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ducted with assistance from regional inspectors to assess the licensee's newly j
implemented Work Planning and Scheduling Section.
Two hundred and fifty hours j
of direct inspection effort were expended for this inspection, i
l Results:
Two violations of NRC requirements were noted during the inspection
j period. One involved inadequate monitoring of site visitors requiring escorts
in the protected area. Another involved improper radiological sampling of the
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drywell atmosphere prior to conducting a vent of the drywell. All other activ-
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ities monitored by the inspector were conducted in compliance with station and
NRC requirements, j
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i 8G11090175 88102G
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PDR ADOCK 05000322 O
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DETAILS 1.
Followup on Previous Inspection Items (MC 93702)
1.1 (Closed) Inspector Followup Items 86-14-01 & 86-14-02:
Engineered Safety Features (ESF) actuations due to Electrical Protection Assem-bly (EPA) breaker trips while Reactor Protection System (RPS) power was being supplied by the alternate feed transformer (86-14-01) &
unplanned automatic actuations of ESF systems caused by power spikes on the grid voltage due to thunderstorms (86-14-02); both items are also referenced in Licensee Event Reperts (LERs)86-029 & -030 and 87-011 & -014. The inspector reviewec'. Independent Safety Engineering Group (ISEG) report 87-026 which decumented the investigation into the cause of the spurious tripping of the EPA breaker. The licensee determined that the cause of the unplanned automatic actuations of the ESF systems during thunderstorms was due to lightning induced voltage spikes which proparpted through to the RPS alternate feed supply and tripped the EPA breaker.
The precise caun of the non-lightning related EPA breaker trips could not be determined but was believed to be due to the low output voltage of the voltage regulator s
installed in the feed circuit (which was only slightly above the
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voltage setpoint of the breaker undervoltage trip).
The licensee
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subsequently installed lightning surge arrestors on the high side of the RPS alternate supply transformer and adjusted the output voltage
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of the voltage regulator upward from 119.7V to 121.5V in an effort to eliminate the spurious EPA breaker trips.
The voltage regulator was also added to the licensee's prev u ative maintenance (PM) program by the creation of procedure PM-SAWS-1C71-410E/E-001-4010.
This pro-cedure requires cleaning of the regulator once every eighteen (18)
months when the EPA breaker surveillance procedure is conducted.
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As a result of the inspector's review, no violations were noted. The
licensee committed to issue a supplement to LER 87-014 to outline the results of ISEG report 87-026 and document the corrective actions undertaken to correct this problem.
These items are closed.
1.2 (Closed) Unresolved item 87-20-01: Damaged rnubbers in the RHR sys-
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tem due to waterhammer events.
The inspector reviewed licensee engineering report NSD-88-1655 dated June 6,1988, concerning the
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failure of snubbers 1E11*PSSP-808 and -810 on the kesidual Heat
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Removal (RHR) System suction line from the Reactor Coolant System (RCS).
The report stated that the probable cause of the damage to
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the snubbers was a waterhammer transient that occurred in the RHR suction piping sometime between March 1986 and February 1987. The
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probable cause of the waterhammer was the unvented drainage of the RHR suction piping followed by the refilling of the line.
The event
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apparently occurred when the reactor was shutdown and the refueling
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- .avity was filled. Suf ficient energy to develop the transient existed at that time due to the hydrostatic pressure exerted on the RHR suction line when the refueling cavity was flooded.
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In response to the event, the licensee replaced the damaged snubbers and made changes to System Operating procedure 23.121.01 "Residual Heat Removal (RHR) System" to preclude the recurrence of waterhammer in this line.
The licensee also completed a previously approved design change that provided for an electrical interlock between the RHR suction valves from the Reactor Coolant System and the suppress-ion pool to prevent a recurrence of a waterhammer event.
The inspector reviewed the NSD-88-1655 report and System Procedure SP 23.121.01, Revision 29. He noted that the NSD report was thorough and appeared to make a concerted effort at finding the root cause of the event, although the precise cause and circumstances surrounding the event could not be determined conclusively. He also found that the changes to SP 23.121.01 should prevent the unvented drainage of the RHR section line and thus prevent a recurrence of a waterhammer in the line.
The inspector physically examined the repairs made to the snubbers as well as the general e 9dition of several additional snubbers and spring can supports on the line.
No problems were noted. This item is closed.
1.3 (Closed) Violation 88-01-01:
Procedural inadequacy pertaining to Local Leak Rate Testin] (LLRT) resulted in overpressurizing the test volume behind Reactor Core Isolation Cooling (RCIC) turbine exhaust isolation valve IE51*MOV-045.
The licensee stated in their response to the violation dated April 18, 1988 that changes were made to pro-cedure SP 84.654.03, Rev.
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"Primary Containant Leak Rate Test" cautioning against overpressurizing test volumes when using water as a testing medium. The caution statement requires the use of a relief valve in the water supply header and recommands the use of a needle or globe valve to control flow into the test volume.
The licensee also performed an engineering evaluation and determined that the valve was not damaged by overpressurizing the test volume.
The inspector for NRC inspection 88-01 reviewed the changes to the procedure and found them acceptable.
This violation is closed.
1.4 (Closed) Violation 87-?2-01: Failure to properly implement a station security procedure.
The inspector reviewed the licensee's responsa to the violation dated April 15,1988 (which was classified as safe-guards information).
The corrective actions stated in the response appear to be adequate to prevent a recurrence of the event.
This violation is close.
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2.
Operational Safety Verification (MC 71707. 71709, 71881, 93702)
2.1 Inspection Activities On a daily basis throughout the report period, inspections were con-ducted to verify that the facility was operated safely and in con-formance with regulatory requirements.
The licensee's management control system was evaluated by direct observation of activities, tours of the facility, interviews and discussions with licensee per-sonnel, independent verification of sarety system status and limiting conditions for operation, and review of facility records. The licen-see's compliance with the radiological protection and security pro-grams was also audited. Significant events which occurred during the inspection period were followed or investigated by the inspector.
These inspections were conducted in accordance with NRC inspection procedures 71707, 71709, 71881 and 93702.
2.2 Inspection Findings The unit remained in cold shutdown throughout the period of this report.
Significant work activities continued to focus on returning systems to operable status to place the plant in a condition that would allow plant operation if desired, in late September. Signifi-cant surveillance activities were focused on completion of 18-month surveillances which would come due by year's end.
These activities were conducted to achieve an operational readiness which would allow uninterrupted power operations from late September through early 1989.
Despite this goal, the licensee curtently has no plans to operate the facility at power.
On July 26, 1988, the licensee perforn:ed a vent of the drywell atmos-phere in violation of the sampling.equirements of the facility Technical Specifications.
Technical Specification 4.11.2.8.2 requires the samples taken of the drywell atmosphere to be analyzed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to commencing a venting operation if the containment purge filter is bypassed.
The venting of the drywell was to be performed to assist in relieving an excessive dif ferential pressure which had existed across the dry-we'.1 personnel access hatch.
Operations personnel notified Radio-chemistry of their desire to vent the drywell via the containment purge filter. In response to this request, Radiochemistry personnel obtained drywell atmosphere samples required by Technical Specifica-tion 4.11.2.1.2-1.
Operations personnel then commenced the vent via the the containment purge filter even though the sample analysis had not been completed. This was consistent with the Technical Jpecifi-cations which did not require the analysis to be completed prior to
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commencing a vent via the containment purge filter. After approxi-mately 15 minutes of venting via this lineup, Operations personnel elected to swap to a higher capacity lineup which bypassed the con-tainment purge filter.
The bypassing of the purge. filter 15 minutes into the vent effectively violated the requirement of Technical Specification 4.11.2.8.2 (88-07-01).
The licensee identified this violation and properly complied with reporting requirements contained in License NPF-36 dnd 10 CFR 50.73.
Additionally, subsequent analysis of the samples taken prior to the vent confirmed that there was no release of radioactive materials.
NRC Inspection Report 50-322/88-06 identified a concern relating to attention to detail in the licensee's effluent monitoring program.
More specifically, this occurrence is the fourth incident resulting in inadequate radiological monitoring of station effluents to have occurred since August 1987.
This violation, even though licensee identified, does not merit use of enforcement discretion to waive a Notic.e of Violation.
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$T!JU5 Sl.TEGOGDS
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'iS lu)T F0!! PULLIC DISCI.0SLt%,,1 B 1 iTEllT10iCLLY LEFI SL ANK.
TlUS l'ARAGRAPH CONTAnts s.mcogy'
INFDEMAl!U!! AND IS HQT F0:t pgnjg DiSOLOSUEE. !T IS INTElill0lluLy LEFT BikHK, The inspector discussed this concern with the licensee management.
This issue will remain unresolved pending NRC management review (88-07-04).
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3.
Surveillance Testing (MC 61726)
3.1 Inspection Activity I
I During this inspection period the inspector performed detailed tech-nical procedure reviews, witnessed in progress surveillance testing, and reviewed completed surveillance packages. The inspector verified
that the surveillance tects were performed in accordance with Tech-t nical Specifications, licensee approved procedures, and NRC regula-
tions. These inspection activities were conducted in accordance with
NRC inspection procedure 61726. The following surveillances were
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SP 44.12'/.24, SP 44.611.03 Recirculation Pump End of Cycle (EOC)
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Trip Response Time Testing;
SP 84.654.03 Leak Rate Test of Main Steam Isolation Valves;
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SP 34.315.03 125VOC Station Battery Weekly Checks;
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SP 34.315.04 125VDC Station Battery Capacity Test;
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SP 24.406.03 Functional Test of MSIV Leakage Control System;
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SP 84.654.04 Orywell to Suppression Pool Bypass Leakage Test.
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3.2 Inspection Findings The inspector noted a discrepancy in the acceptance data for the Station Battery Capacity Mic.
The acceptance criteria required the
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battery to discharge greater than 96 amperes throughout a 119 minute l
period. The data point recorded at the 75 minute interval was 95
amperes.
All other data points were acceptable. Although the test L
results had not been completely reviewed by the licensee, the battery e
had been declared operable based on a preliminary review and approval
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of the data.
The licensee determined that the discrepancy negated i
the results of the test.
The battery was declared inoperable and I
efforts were initiated to reperform the surveillance.
No viola-I tions of facility Technical Specifications resulted since the limit-i ing Condition for Operation in cold shutdown was met without the
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battery being operable.
No other violations or inadequacies were (
noted.
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The licensee completed a significant number of low frequency surveil-
. lances during this inspection perf od. The majority of these surveil-lances were conducted by the Instrument and Controls (I&C) section.
Generally, these low frequency I&C surveillances tend to be more complicated than routine surveillances and, during previous periods, have resulted in an increased frequency of inadvertent Engineered Safety Features (ESF) actuations.
The inspector noted that no
Although two inadvertent ESF actuations occurred during this period, these were not related to the increased I&C surveillance activity.
Licensee Event Report (LER) 88-12 documents an ESF actuation which occurred as a result of routine I&C maintenance on a water level instrument.
Also, an inadvertent ESF actuation occurred on September 7,1988 during the performance of the drywell to suppress-ion pool bypass leakage test.
The cause of this actuatien '.tas not directly related to the surveillance but was the result o/ an opera-tor inadvertently restoring a system to normal status Ming perform-ance of the test.
The inspector also noted a low failure rate during the performance of these low frequency surveillances. This would indicate a success-ful preventive maintenance program resulting in a high degree of facility readiness for power operations. For example, the licensee had anticipated the Main $+.eam Isolation Valves (MSIV) might fail the leak rate testing resulting in extensive rework. This assunption wa*
based on industry experience.
The MSIV's passed the leak rate tests without any rework. The failure of the battery test noted above was the result of technician error and was not considered significant to this assessment.
4.
Maintenance Activities (MC 62703)
4.1 Inspection Activity During this inspection period the inspector observed selected main-tenance activities on safety related equipment to ascertain that these activities were conducted in accordance with approved proced-ures, Technical S..ecifications, and appropriate industrial codes and standards.
These inspt.ctions were conducted in accordance with NRC inspection procedure 62703.
The following activities were reviewed by the inspector on an ongoing basis:
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SP 34.315.04,125VDC, Station Battery "B" Replacement;
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MWR's 88-2709 & 88-1762, Inspect and Rework valves E41%1-042
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and E41*MOV-043; MWR 88-2569, Repair of Service Water Strainer; and,
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Repair activity associated with station ventilation exhaust fans.
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4.2 Inspection Findings No unacceptable conditions or violations were noted.
5.
Engineered Safety Feature (ESF) System Walkdown (MC 71710)
5.1 Inspection Activity
The inspector verified the operability of selected ESF syr,tems by
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performing a walkdown of accessible portions of the system to confirm that system lineup procedures match plant drawings and ' the as-built
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configuration. This ESF system walkdown was also conducted to iden-
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tify equipment conditions that might degrade performance, to deter-i mine that instrumentaticn is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriate.
i The Reactor Building Closed loop Cooling Water system and selected portions of Emergency Core Cooling systems were examined.in detail.
This inspection was conducted in accordance with NRC inspection pro-
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cedure 71710.
5.2 Inspection Findings i
No unacceptable conditions or violations were noted.
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6.
In Office Review of Licensee Event Reports I
The inspector reviewed Licensee Event Reports (LERs) submitted to the NRC i
to verify that details were clearly reported, including accuracy of the
cause description and adequacy of corrective actions.
The inspector
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determined whether information was required from the licensee, whether
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gene-ic implications were involved, and whether the event warranted onsite follow-up. These reviews were conducted in accordance with NRC inspection procedure 92700. The following LERs were reviewed:
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LER 88-11:
Primary containment purge bypassing containment purge filter without obtaining sample analysis. See paragraph 2.2.
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LER 88-12:
ESF actuation resulting from maintenance on water level
instrumentation.
The inspector had no further questions related to these LERs.
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7.
Inspection of Work Planning and Scheduling Section
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An inspection was conducted to assess the effectiveness of the licensee's i
newly implemented Work Planning and Scheduling Section (WPS). The Itcensee implemented the section in December, 1987 in response to a perception that more efficient coordination of station work activities was needed. This r.aed was identified by the licensee but was clso an NRC concern (see NRC Reports 50-322/86-99 and 87-20).
The purpose of the WPS it to coordinate preventive maintenance, corrective maintenance, and surveillance activ-ities among the various plant staff sections in accordance with station goals and facility requirements. Personnel from each section are assigned to the WPS on a temporary, rotating basis. Generally, these persons are
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highly experienced foremen or supervisors and are responsible for planning
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activities within their area of expertise.
These work planners are in
close proximity to each other and are overseen by two Lead Planners who maintain active Senior Operator licenses. Each planner works closely with his associated section and inputs required work activities to the WPS.
Formal schedule meetings are held on a weekly basis to coordinate ongoing and future work activities. Both routine and emergent work is prioritized and scheduled according to overall station goals.
This has resulted in more efficiert management of sta* ion resources and has allowed greater control and monitoring of station activities by senior management.
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Previously, each section was responsible for coordinating its own work activities. A lack of centralized planning resulted in many inefficien-cies.
For example, a component would be removed for preventive mainten-ance to be perfomed by the mechanical maintenance section. Following the i
maintenance, operations would restore the component to normal status and perform a surveillance to prove the component operable.
In many in-
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stances, following restoration of the component, other sections would request to perform maintenance on the same component.
The WPS has successfully eliminated this type of inefficiency.
j The WPS has also taken lead responsibility in controlling the Maintenance
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Work Request form program.
These forms are used to request, describe, i
authorize, and document work activities.
Prior to the WPS, tne backlog i
of open MWR's was approximately 2600. The WPS conducted an extensive pro-l gram to va'idate these outstanding KdR's.
Through this procoss it was found that many of these MdR's were duplicates, were no longer applicable i
due to station modification, or were simply held up in the review process.
An aggressive program to reduce this backlog of unnecessary MdR's was implemented.
The WPS also implemented a policy of "walking down" all new MWR's prior to serialization to verify the deficiency and eliminate
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These activities have resulted in a backlog reduc-tion of outstanding MdR's to approximately 1700.
Many of these MdR's
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still outstanding are related to the Colt Diesel modification. These will remain outstanding until completion of the modification during the first refueling outage.
Also, a large number of KdR's were initiated in the last six months as a result of several large inspection programs.
The inspector noted that there was still a small number of K4R's which were quite old and could be closed.
The licensee intends to complete these relatively low priorty activities as resources permit. Considering these factors, the licensee has sv.cessfully reduced the KdR backlog to a more reasonable and manageable number consistent with station goals.
Outstanding MdR's are also tracked and monito ed in a more efficient and meaningful manner.
Much of the program has been computerized and the licensee intends to eventually computerize the entire KdR program. Fre-quent (weekly) reports are prepared which trend several parameters asso-ciateo 11th the K4R program.
The inspector found these reports to be useful in analyzing the program effectiveness.
For example, through a brief review of the report, one can determine changes in the backlog of open K4R's by section, system, or work stage.
These reports are dis-tributed to senior management as well as to each individual section.
The WPS has also helped to reduce administrative burdens on section per-sonnel.
Work planners are responsible for preparing the work packages prior to release of the work to the individual section.
This includes obtaining equipment hold-of f s (i.e.
tagouts), work procedures, component drawings, materials and authorizations. Previously, these activities were performed by a section foreman thus reducing the available time this person could spend in the field.
The inspector found that most foremen now spend a greater time in the field as a direct result of this reduction in administrative burden.
The inspector found that improvement in this program could be gained by better coordination with the warehouse. Currently, there is no represen-tative from the warehouse in the WPS.
Spare parts and their availability are crucial in coordinat'on of maintenance activities.
The inspector found no instances where lack of this coordination resulted in challenges to safety; however, it was suggested to the licensee as an area for improvement.
In summary, the program has been well implemented and appears to be func-tioning adequately.
Improvement could be gained in further direct involvement of other plant sections and in further automation of the process.
Overall, implementat:on of the program represents a vast improvement over previous practic _ _.__ _ _ _
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Ultimate Heat Sink During Extended Hot Weather The facility's ultimate heat sink (Long Island Sound) did not exceed the maximum design temperature specified in the Updated Safety Analysis Report during the inspection period.
This was a concern at other facilities throughout the country due to the extreme heat and lack of rain during the past summer.
Shoreham d es not have a temperature limitation on the ultimate heat sink speci 'ied in the facility's Technical Specifications.
Section 9.2.1.2 of the Updated Safety Analysis Report specifies a Service
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Water design inlet temperature of 77 degrees F.
During the month of August, intake canal temperatures reached a maximum 75 degrees F but mostly remained a few degrees below this maximum. This was not a concern to the licensee, particularly since the facility remained in cold shut-down.
The inspector did not identify any safety significant problems attributable to the extended hot weather.
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Concrete Block Wall Cracking
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The licensee has discovered stress cracking in various concrete block walls throughout the f acility.
These were found by the licensee during inspections which were conducted in response to NRC Information Notice 88-67, "Lessons Learned From Regional Inspections of Licensee Actions in Response to IE Bulletin 80-11."
Although the cracking appears to be relatively minor, the licensee is evaluating each crack to determine its impact on seismic qualification assumptions.
The inspector will follow the licensee's corrective actions related to this problem (88-07-03).
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Management Meetings (MC 30702)
At periodic intervals during the course of this inspection, meetings were held with the licensee management to discuss the scope and findings of
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this inspection. An exit meeting was held on September 27,1988 to dis-l cuss the preliminary findings of this report.
The inspector also attended entrance and exit interviews for inspections
conducted by region-based inspectors during the period.
l These activities were conducted in accordance with NRC inspection proced-ure 30702.
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