IR 05000443/1987023
| ML20236P565 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/10/1987 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236P546 | List: |
| References | |
| 50-443-87-23, IEIN-86-105, NUDOCS 8711180048 | |
| Download: ML20236P565 (12) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ - - _ _. _ , . . U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-443/87-23 Docket No.
50-443 License No.
NPF-56 Permit No.
CPPR-135 Priority Category B/C -- Licensee: Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit 1 Inspection At: Seabrook, New Hampshire Inspection Conducted: September-9 - October 19, 1987 Inspectors: A. C. Cerne, Senior Resident Inspector D. G. Ruscitto, Resident Inspector Approved By: <[ [ ~ II[/o/f7 h DonaldR.Haverkamp, Chief,ReacprProjects Date Section No. 3C Inspection Summary: Inspection on September 9 - Oct_ober 19, 1987 (Report No.
50-443/87-23) I l Ar_aas Inspected: Routine inspection by two resident inspectors (115 hours) of work activities, procedures, and records relative to startup testing and license issuance; maintenance, surveillance and plant operations during cold shutdown conditions; and licensee event reports.
The inspectors also reviewed licensee actions on a previously identified item and performed plant inspection tours.
Results: No violations were identified.
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, . DETAILS 1.
Persons Contacted T..C. Feigenbaum, Vice President, Engineering and Quality Programs W. J. Hall', Regulatory Services Manager D. E. Moody, Station Manager.
G. S. Thomas, Vice President, Nuclear Production J. M. Vargas, Manager of Engineering J. J. Warnock, Nuclear Quality Manager Interviews and discussions with other members of licensee and contractor management, and with their staffs,.were also conducted relative to the inspection of items documented in this report.
2.
Plant Status During this. reporting period, the plant remained in operational Mode 5, cold shutdown, with primary temperature about 115 degrees F and depressurized.
j-Several events of minor safety significance occurred during this inspec-tion period. These events are documented below: a.
September 5, 1987 - 345 KV Bus Failure. This failure was initially reported in NRC inspection report (IR) 443/87-16.
The "A" diesel generator was out of service for maintenance as allowed by the tech-nical specifications when this event occurred. The fault was deter-mined to be in an SF6 bus duct on the incoming Newington line 369.
Due to plant conditions at the time of the line loss and the con-tinued availability of the Scobie line 363, this occurrence had no ' safety implications.
Evaluation of lessons learned and detailed analysis of equipment response will be the subject of NRC inspection after issuance of the station information report.
b.
September 23,1987 - Inadvertent Engineered Safety Features Actuation Signal (ESFAS). A turbine trip /feedwater isolation signal (P-14) was generated during work on electrical leads associated with level transmitter FW-LT-518 for "A" steam generator.
The leads for FW-LT-519, also on "A" steam generator were incorrectly lifted making up the 2/4 coincidence for P-14.
No safeguards actuation occurred because of plant status at the time.
The inspector concurred with the licensee evaluation that this ESFAS was not reportable in accord-ance with 10 CFR 50.72 or 50.73. Further discussion of this analysis of inadvertent actuations is in paragraph 10 of this report.
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. Inadvertent Reactor Trip Signal (RTS).
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October 5, 1987 - signal,. generated from Lo-Lo level in "D" steam generator, occurred i during installation of the new Rosemount transmitter on FW-LT-548.
! This channel was in.a test-condition when a low spike was received on FW-LT-547 satisfying the 2/4 coincidence for reactor trip.
The reactor trip breakers did not cycle since the solid state protection system was in test at the time. The cause of the spike has not been determined as yet and will be evaluated when the station information report is completed.
Further discussion of analysis of inadvertent actuations is in paragraph 10 of this report.
3.
plant Inspection Tours The inspectors observed work activities in progress, completed work and plant status in several areas during general inspections of the plant.
The inspectors examined ' work for any apparent defects or noncompliance ) with regulatory requirements or license conditions.
Particular note was i taken of the presence of quality control inspectors and quality control l evidence such as inspection records, material identification, nonconform-ing material identification, housekeeping and equipment preservation. The inspectors interviewed station staff, craft, quality inspection and supervisory personnel as such personnel were available in the work areas.
a.
The inspector examined certain field installations, witnessed ongoing l maintenance activities, and checked specific valve line-ups as noted .below: -- Fire wrapping, in accordance with the one-hour, fire-rated bar-i ' rier requirements of the Seabrook Station Fire Protection (Appendix R) Program, was examined and discussed with engineer-ing personnel.
The configuration of newly installed steam generator level l -- transmitters inside containment was checked and discussed with maintenance department I&C personnel.
The in progress repacking of containment building spray valve . -- CBS-V-53 was witnessed and discussed with the maintenance craft I and QC inspector at the work location.
l The position of valves CS-HCV-128, and RH-V-18 and 19 was field -- checked and discussed with operators on shift, with regard to the status of residual heat removal letdown flow to the chemical and volume control system.
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. b' During control room observation periods, the inspector reviewed con- ) , trol room. logs and records including night orders, shift journals, shift turnover sheets, completed repetitive task sheets, the tempor-ary modifications log, weekly surveillance schedules. and ' control board indications.
Specific note was taken of equipment in " pull-to lock" conditions, equipment tagged, alarm status and adherence to i technical speci.fication limiting conditions for operation and action statements.
J With respect to all of the above plant. inspection tour and independent l inspection items, no violations were identified.
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Licensee Actions on a Previously Identified Item (Closed) Unresolved item 87-02-01: Loss of Power to Vital Instrument Panel. As described in inspection report (IR) 443/87-02, an engineered safety features (ESF) actuation occurred upon loss of the uninterruptible power supply '(UPS) from the inverter EDE-I-1E.
While searching for a ground fault, the a-c power supply breaker from motor control center MCC E-512 to EDE-I-1E was opened. The supply of d-c power to inverter IE was available, but the 1nverter output to the 1E vital instrument panel-EDE-PP-1E was delayed for two seconds, causing multiple ESF actuations.
,Upon signal initiation, all systems functioned as designed.
' -Licensee event report 87-006 was submitted by letter (NYN 87038), dated March 23,1987 and indicated that the probable cause of the problem was a fault protective circuit within the Elgar inverter.
This circuitry pro-tects the inverter from ground faults by interrupting the output power at two-second intervals until the fault is cleared. The inspector discussed with the lead electrical system support engineer'the probable cause of.the problem, the generic impact upon the operability of both Elgar inverters 1E and 1F and the planned testing which would be conducted to confirm and correct the deficiency.
Performance testing on inverter 1E was conducted l in accordance with special procedure ES 87-1-32 written to troubleshoot the observed problem.
The inspector confirmed that a 10CFR50.59 evalua-tion was completed in conjunction with the subject procedural approval.
Work request (WR) 87W2706 was initiated to commence testing on inverters 1E and 1F.
Supplemental WR 87W2924 was implemented to augment the test-ing.
It was determined that a transducer card in the inverter logic cir-cuitry to the fault protection function was unduly sensitive to noise induced by faults elsewhere in the electrical supply system.
The exist-i ence of the initial ground fault on 480 volt Bus E51 adversely affected the ability to transfer to the UPS DC supply, because the inverter cir-cuitry sensed the continual presence of a fault.
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Elgar, the inverter vendor,.was involved in the testing and troubleshoot- 'ing activities and concurred with the analysis that pinpointed the problem to be.with the sensitivity-of the inverter transducer cards. Elgar engi- . neering change notice (ECN) 6384 was issued to replace several resistors on the subject transducer cards with ones having one-tenth _ the original j resistance,- to reduce the operating gain level and thus eliminate the .
undesirable noise. This design change was implemented with New Hampshire Yankee : design coordination report (DCR) 87-147 and work was completed under WRs 87W3453 and 3454.
Subsequent testing to the requirements of.
ES87-1-32 provided evidence that the problem had been corrected and that d the acceptance. criteria with respect to UPS output for the. observed AC i fault conditions had been met.
The inspector reviewed the subject DCR, ECN, WRs and ES test results and
discussed completion of the modifications with technical support, opera-i tions and licensing personnel.
The inspector observed that licensee planned troubleshooting extended beyond the identified problem (the E51 bus fault impacting inverter IE) to a generic review of the inverter design as the problem also affected inverter 1F. The inspector also noted Yankee Atomic Electric Company (YAEC) involvement in the independent review of the DCR.
The corrective action initiated by the licensee in response to the event identified by LER 87-006 is complete and the ques- - tioned operability of the subject inverters has been resolved with the noted component design modifications.
This item is closed.
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Licensee Event Reports ! a.
(Closed) LER 87-006: Loss of Power to a Vital Instrument Panel. See paragraph 4a for discussion and closure of this LER in relation to a previous NRC unresolved technical concern, b.
(Closed) LER 87-014: Inadvertent Diesel Generator Start. As docu-mented in IR 443/87-10, paragraph 2g, the trip of the 'F~" train 4.16 B kV vital bus E6 resulted in the starting and loading of emergency diesel generator (EDG) IB.
As was noted, upon initiation of this event, all systems operated as designed. Since testing of one of the second level undervoltage protection relays (620) was in progress when this event occurred, faulty test equipment was initially sus-pected.
Subsequent licensee evaluation has resulted in the identi-fication of a deficiency in the surveillance testing procedure to be the cause of this spurious ESF actuation.
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Thef 62Dj relay Lis? designed to provide Ja! ten-second delay-between the ~ time when; undervoltage on. the 4.- 16 kV bus' is. sensed.and. the-point. at' ' ' which.the emergency' power _ sequencer is actuated.
This' actuation sequence starts the EDG, opens the offsite' power supply _ breakers, and; controls EDG loading. This timed. delay allows for spurious-voltage transients which may be ' caused by large. load starts.
In accordance-with; surveillance procedure, MX0513.07, the 62D relay test-isa con - ducted utilizing a test configuration which:appli'es_d-c control power.
through a test' switch to the relay and measures' the relay response , time on the component delay timer when the d-c power is removed.
However, when Lthe ; timer was ' reset in this test configuration,. the-resulting resistance drop was sensed by the logic circuitry as timer run-out with subsequent pick-up of the 62D relay. The _ designed ESF actuation followed as'a conseq~uence.
To preclude recurrence of' this event, the surveillance procedure was revised to 'open additional contacts in the emergency power' sequencer prior to resetting the timer at the start of the relay response time measurement.
The. inspector reviewed the revisions to' the affected . procedural steps (Change No.2 to MX0513.07, Revision 1) and noted
'that they effectively precluded the 620 -relay testing process from erroneously causing sequencer actuation. The inspector also reviewed .LER 87-014, submitted by letter NYN-87073, dated May 29,1987 and determined,that corrective action, as stated, had been completed and ~ that reporting requirements in accordance with 10CFR50.73 had been me t.' This LER is closed.
6.
10 CFR 21 Reports a.
~(Closed) 10 CFR.21 Report 87-88-02: Common Mode Failure of Containment Equipment Hatch Air Lock Door Equalizing Valve Linkages.
LER 87-004 was issued in relation to this component deficiency and closed in IR 443/87-10.
Unresolved item 87-08-05 was also opened on this subject and subsequently closed in IR 443/87-22. The 10 CFR 21 report was issued by licensee letter NYN-87059 dated April 24,1987 and discussed the existing defect, a worst-case analysis of post-accident consequences, and the corrective action in the form of l component design modifications.
Additionally, NUREG/BR-0051 " Power
Reactor Events", (Volume 9, No.1 section 1.2), discussed the problem in the overall context of the unusual event that resulted from the subject failure.
L During this inspection, the inspector reviewed DCR 87-106, which redesigned the equalizing valve linkage assemblies for the air lock L doors which are an integral part of the containment equipment hatch.
The inspector noted that implementation of this design change super-seded the short-term modifications (DCR 87-073 and 075) implemented soon after the occurrence of the unusual event in February, 1987.
The inspector reviewed the final design, witnessed a demonstration of manual door operation, and discussed the changes with the responsible technical support engineering personnel.
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J { l ( The inspector noted that of the twelve original pin / bushing sub - i assemblies (six per. door) which were susceptible to the subject link- -{ age failures,. only four had the capability to adversely affect more
than one door equalizing valve. The event of February 11, 1987 found two of these four " worst case" links- (one per door) disengaged, .resulting in all four equalizing valves. being open.
The licensee corrective action, however, has addressed all twelve common link connections, installing a machine screw into ' the modified, drilled and tapped pin, with a new plate washer preventing the pin from sliding. through its link. Thus, the new design effectively precludes recurrence of the previous common mode failure.
The inspector also noted that local leak rate testing in accordance with surveillance procedure EX1803.003 was conducted during this inspection period. The full barrel test on the equipment hatch per- -sonnel air lock,'as required by TS 4.6.1.3.6 once every six months or af ter the completion of maintenance, resulted in a leak rate within acceptable design criteria.
The inspector reviewed the test results to verify that the operability of this air. lock had not been adversely affected by the design modifications to the manual operating linkages.
The corrective actions documented in this 10 CFR 21 report have been completed and the air lock retested.
The specific component defect resulting in the common mode failure of air lock doors has been fixed.
The inspector has no further questions on.this issue. This item is closed.
b.
-(0 pen)
CFR
Report (87-88-03): Service Water Valve Seat Problems.
Licensee efforts continued in the repair of service water system butterfly valves. The "A" train valves were repaired and the system restored to service.
In conjunction with inspection and 'l repairs to the "B" primary component. cooling water heat exchanger, "B" train valve repairs were started. The inspector monitored valve repair progress and observed removal, disassembly, repair, testing and restoration of selected valves. No safety concerns or violations were identified. The completion of licensee corrective actions will be followed up in future NRC inspection.
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(0 pen) 10 CFR 21 Report (87-88-04): Gould Relay Failures. In April ! and August, 1987 relays in the condenser water box vacuum breaker i valve position circuit failed to actuate.
Licensee investigation of ! this problem revealed that these normally energized relays had over-heated.
Samples of the failed relays were sent to the manufacturer,
Telemecanique, for analysis.
The vendor evaluation of the failure mode indicated that the special relay coil employed for the specific . l L
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Seabrook application could withstand. the service operating tempera-ture but the armature carrier could not. Telemecanique notified the NRC on October 9,1987 of this defect in accordance with 10 CFR 21.
A licensee investigation is in progress to determine the number and use of the relays installed at Seabrook. This item will be adminis-tratively tracked under open item number 87-88-04 and will be followed up in future NRC inspection.
7.- Security Issues The inspector discussed the following items with the NHY security depart-i ment supervisor and with other station and engineering personnel, as i necessary to resolve certain technical and administrative questions of a ~ safeguards nature: a.
Vital area controls and the designation of certain plant areas as vital in accordance with the NHY Physical Security Plan, b.
THe DCR 87-0080 implementation and the associated engineering evalu-ation relative to access controls and security / operations contingency l options during emergency conditions.
c.
Visitor escort controls and security actions in response to SIR 87-90.
d.
Key controls and 10CFR73 deportability relative to SIR 87-087.
. e.
The NHY administration of the Fitness for Duty Program and the deportability of certain incidents involving controlled substances with respect to NRC guidance provided by Regulatory Guide (RG) 5.62.
The specific details of certain of the above inspection issues constitute safeguards information.
The NRC follow-up activities have included a review of design changes and other engineering information, an evaluation of the coordination between the security and operations departments to ef fect safe-shutdown contingency controls, and discussions with the NRC Office of Nuclear Reactor Regulation (NRR) safeguards branch reviewers ! regarding the definition of vital areas (Review Guideline No.17, Revision 1).
The inspector determined that the existing security controls ade-quately addressed the situations encountered and the questions raised in the above areas.
Discussion with the security department supervisor revealed that additional programmatic enhancements may be forthcoming particularly with respect to security contingency planning for outages, for deportability and for the response to certain types of incidents.
The inspectors will continue to monitor the physical security program imple-mentation both routinely and in response to incidents which require j compensatory security measures.
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l With respect to'the. NHY Fitness for Duty Program, the ' inspector reviewed NHY. policy. directives 10.0 through '10.4 and evaluated NHY security response.to an incident on September 14, 1987 involving 'the detection and suspected presence of. an unauthorized, controlled substance on site.
Security department search techniques were implemented at.the time with negative results.
Personnel accountability was tracked through the security key card system and chemical screening in accordance with both Directive 10.3 and-the Station Security Manual was administered to certain station. personnel. The results of the chemical screening process iden-tified a plant health physics technician who required referral to the NHY Employee ' Assistance Program.
This individual's access to the protected area was suspended.
While licensee actions in response to this incident did not substantiate the existence of a controlled substance' on site, they did evidence thorough followup of a suspected ' problem.
The inspector confirmed that .the applicable fitness for duty directives had been followed and noted, in the given example, that the program was effective in identifying a poten-tial personnel problem. No unresolved safety questions were identified as a result of the subject incident.
In regard to all of the above security issues, no safeguards concerns or violations were identified.
8.
Independent Inspection a.
IE Information Notice 86-105. This information notice (IN) described the potential for loss of intermediate power level reactor trips due to a failure of the P-10 interlock.
The inspector reviewed abnormal operating procedure OS1211.04, " Power Range NI Instrument Failure", and major plant evolution procedure OS1000.03, " Shutdown from Minimum Load to Hot Standby".
The inspector determined that the concerns discussed in IN 86-105 were adequately addressed to ensure appropri-ate operator action should a nuclear instrument fail during plant shutdown, with another instrument channel already in the test condi-tion. No safety concerns or violations were identified, b.
Emergency Diesel Generator Protective Trips.
The inspector reviewed the Seabrook Station Final Safety Analysis Report (FSAR), training department system descriptions, and NHY logic
diagrams and electrical schematics to determine the type and number of protective trips provided for the emergency diesel generators (E0Gs) by design and under what conditions these trips would be by-passed.
These protective trips can be divided into two general groups: w_____.
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_-_ _ p 4 dc 9-l I' L(1)' EDG trips bypassed with a safety injection signal.("S") (a) High jacket cooling water temperature (b) High lube oil temperature (c) Energization of the back-up protective lock-out relay (86DB) which protects against certain EDG problems, e.g., reverse power, in relation to the 4.16 kV bus (2) EDG trips which are not bypassed with an "S" signal: (a) Low lubricating oil pressure (2/3 logic) (b) Mechanical overspeed of the engine (c) Energization of the primary protective lock-out relay (86 DP) which protects against differential phase current-faults Another protective function is provided by the bus protective lock-out relay (868) which protects against a 4.16 kV bus fault. This relay does not trip the EDG, but does open the EDG output breaker and is not bypassed when an "S" signal is present.
These protective functions have been designed in compliance with the NRC Branch Technical Position (BTP) ICSB 17. The inspector evaluated the " Diesel Shut-Down and Start Logic Diagram" (1-NHY-503492) for logic and interlock development consistent with the FSAR discussion on diesel trips. The inspector also reviewed the surveillance pro-cedure 0X1426.02 for the 18-month operability criteria of EDG 1A and noted that the EDG lock-out features are tested. The inspector ver-ified that the "S" signal bypass of certain EDG trips are consistent with the interlocks listed in subparagraphs (1) and (2) above.
While loss of offsite power (LOP) testing is also conducted as part of procedure 0X1426.02, no EDG protective functions are bypassed directly as a result of a simulated LOP without a coincident "S" signal.
Some of the logic development associated with energization of the 86DB relay may routinely bypass the protective trip upon a LOP. However, this is the case only because the subject backup pro-tection requires the incoming supply breaker to the 4.16 kV bus to be closed.
Upon a LOP the emergency power sequencer activation opens these supply breakers, thereby preventing this particular 86DB trip as a coincidental result, but not as a specific design requirement of the LOP event itself.
Therefore, while no specific EDG protective ! trips are bypassed by a LOP, unless an "S" signal is also present, i breaker positioning in relation to the sequencer operation may affect ) certain protective functions.
The inspector examined the EDG pro-l tective trip design and interlock features for any cases where non-essential trips might defeat EDG safety functions.
None were iden-tified.
Review of the logic and schematic drawings revealed a design I consistent with FSAR commitments and system descriptions.
No safety concerns or violations were identified.
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Station' Information Report 87-070: Unit Substation' 12 Breaker
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10verheating.
During performance of-routine maintenance a current > transformer ~ in; the1 secondary breaker compartment of. unit substation 12 : (ED-US-12) = was "found to have ~ overheated. Licensee investigation revealed an L undersized hole drilled-in the -buswork causing a loose ~ connection.
The inspecto'r reviewed the ' SIR and ~ work; request for ' . repairs.and had no: questions.
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Maintenance ' 'a.
Primary Component' Cooling Water System-The l inspector observed field weld repair of the pinhole leak on valve ' CC-V-298, the' "D" primary component cooling. water (PCCW); pump dis-charge. check valve... The inspector reviewed work package ' 87W004556 which.in'cluded the ASME Section XI travellers.
! . The : inspector-' also observed eddy. current testing (ECT). of the "B"' ' train ~PCCW heat. exchanger CC-E-178. Tube sleeving was commenced when-the ECT results indicated tube degradation similar to that found on ~ 'the "A" train,- as was ' discussed in IR 87-16, Section' 3a.
The inspector discussed the process with the vendor representative and also.noted the presence of a NHY QC inspector.
q The inspector reviewed SIR 87-076 which discussed damage to the "A" H train' PCCW heat exchanger CC-E-17A.
.This. report contained NHY j Engineering - Evaluation 87-001 concerning the. repairs to CC-E-17A.
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- The inspector had no questions. The repairs to 'CC-E-178 will be - the.
! subject-- of continued NRC follow-up inspection as repairs progress, ' b.
Service Water System As part of the inspection associated with repair of service water (SW) system valve seats and the PCCW heat exchangers, the inspector entered the SW inlet piping to the' "A" PCCW heat exchanger CC-E-17A to examine a cement-lined pipe joint repair that evidenced minor flaking near the inlet flange.
In response to the inspector's , inquiry, NHY personnel produced non-conformance report NCR 4837 i written in August, 1983 by Pullman-Higgins, the piping supplier.
The . repair was properly documented and conducted in accordance with pro-cedures IX-30 and IX-31.
Further discussion with the NHY systems support manager revealed that several sections of SW piping will be inspected during periodic entry for routine maintenance such as strainer cleaning.
To date, the Belzona repair of the pipe lining
, p has shown no noticeable degradation although some joints, such as the j one repaired above, are exhibiting minor flaking.
The above joint was repaired and the inspector examined photographs of other SW internal piping sections taking particular note of repair joints. No safety concerns or violations were identified.
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. < 110.;' Inadvertent' SSPS.and ESFAS Actuations Section? 2. ofl this report included discussions of.'two events related 'to inadvertent actuations ' of the SSPS and the; ESFAS, As a result of the - shutdown ; plant ' condition, the actual individual consequences of these events are - of ~ no '.' safety significance, and neither was-reportable in-accordance with 10..CFR 50.72 or 10 CFR 50.73. The' actuations are', - how-. . ever,;the latest in a series of sixteen inadvertent' actuations over the.
last year. ' The list below documents only those additional items not con-sidered ' reportable. Those items meeting 'the g reporting.. requirements of-10 CFR.50 : have all been documented in previous ENRC inspection reports.
This list-also specifically ' excludes those events' primarily or wholly caused. by equipment failure.
Only these incidents with' operational 'causes, -1.e., operator 'or procedural errors _ are li sted.
, '01-15-87' Inadvertent FWI.
SIR 87-007 04-1-87 Inadvertent FWI.
SIR 87-034 04-7-87 Inadvertent RTS.
SIR 87-040 07-9-87 Inadvertent FWI.
SIR 87-062.
09-23-87 ' Inadvertent FWI.
See Section 2 10-5-87 Inadvertent RTS.
See Section 2 The inspectors discussed. the. above actuations with NHY' staff members at the=e'xit meeting conducted-on October 21, 1987. The. licensee subsequently committed to the formation of an internal task team to evaluate the events forLroot cause in order to determine if any salient connections exist.
The NRC will track the progress of the licensee's further evaluation of these events and any future actuations.to determine whether.some program-matic deficiency might have contributed to the cause of these events, and to assess the adequacy of licensee corrective measures.
(0 pen item 87-23-01) 11. Open Items
.0 pen items are matters that require further review and evaluation by the inspector.
Open items are used to document, track and ensure adequate' j followup on matters of concern to the inspector.
' 12. Management Meetings At periodic intervals during the course of this inspection, meetings were held with plant management to discuss the scope and findings of this inspection.
An exit meeting was conducted on October 21, 1987 to discuss the inspection findings during the period.
During this inspection, the ' NRC inspectors received no comments from the licensee that any of their inspection items or issues contained proprietary information. No written material was provided to the licensee during this inspection.
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