IR 05000213/1987018

Resident Insp Rept 50-213/87-18 on 870610-0714.No Violations Identified.Major Areas Inspected:Plant Operations,Radiation Protection,Fire Protection,Security Maint & Surveillance. Five Unresolved Items Closed

ML20236J591
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/28/1987
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20236J567	List: IR 05000213/1987018 ML20236J564
References
50-213-87-18, NUDOCS 8708060215
Download: ML20236J591 (12)

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U.S. NUCLEAR REGULATORY COMMISSION  !

, REGION I Report No. . 50-213/87-18 l

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.-Docket N License No'. OPR-61-Licensee: Connecticut Yankee Atomic Power Company P.-0. Box 270 ,

Hartford, CT 06101 j Facility: Haddam Neck Plant, Hadvam, Connecticut ,

Inspection at: Haddam Neck Plant, Inspection dates: June 10, 1987 to July 14, 1987 i Inspectors: Andra A. Asars', Resident Inspector Geoffrey E. Grant, Resident Inspector, Millstone 1&2 William J. Raymond, Senior Resident Inspector, Vermont Yankee '1 Paul D. Swetland, Senior Resident Inspector  !

Approved by: AO <

7 bb I E. C. McCabe, Chief, Reactor Projects Section 3B Date .

Summary: Inspection 50-213/87-18 (6/10/87 - 7/14/87) l l

Areas Inspected: This was a routine resident inspection (204 hours0.00236 days <br />0.0567 hours <br />3.373016e-4 weeks <br />7.7622e-5 months <br />). Areas re-viewed included plant operations, radiation protection, fire protection,' security, I maintenance, surveillance,. events occurring 6uring the inspection period, open j items, switchgear building construction, environmental qualification of cable con-nectors, and a potential generic issue concerning the Loss of Load / Turbine Trip- l Analysi j Results: No violations were identified. No new items were opened. An open in- .

spection item concerning Engineering Expertise on Shift (see detail 5.4) requires )

further licensee action before closure. Five unresolved items relating to Ultra- ]

sonic Examination results interpretation (see detail 5.1), Emergency Diesel Genera- j tor jacking practices (see detail 5.2), Emergency Plan Implementing Procedure re- ,

visions (see detail 5.3), surveillance procedures for masonry walls (see detail ]

5.5), and repair procedures for masonry walls (see detail 5.6) were close '

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TA!RE OF CONTENTS

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. , Summa ry of Faci l ity he ti vi ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .J ........ 1 , Review of Plant Operations........................................... 1 Plant Operations Review Committee.................................... 2 )\ Observation of. Maintenance and Surveillance

, i Activities............... 2 1 ) Followup on Previous Inspection Findings...... >

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5.1 Ultrasonic Examinatica Results InterpretaWm, . t

5.2 Review of Emergency Diesel Generator JackiQ bactices. . .:. . . . . . . 2

, 5. 3 Emergency Plan implementing Procedure Revis1pns....'.... .

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5. 4 Engineering Expertise on Shift..............'..'.'.i............... 3

' 5. 5 Surveillance Procedure for Masonry Block Walls. 0............... 3 5.6 Specification Procedure for Repair of Masonry Block Walls....... 4 Followup on Events Occurring During the Inspection......q............ 4

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S.1 Licensee Event Reports.......................................... 4 (. ,l '

6.2 Empl oyee Found Wi th Mari j uana. . . . . . . . . . , . . . . . . . . . . . >. . . . . . . . . . . . . 5 6. 3 Primary Water Storage Tank Level Transmitter Line broken. . . . . , .

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6.4 Dropped Control Rod and Turbine Runback.............. ...... ... 6

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6. 5 Inoperable Rod Position Indicator........ ...................... 6

, Review of, Periodic and Special Reports...............................

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! # Cable f.cdectov QuaMfication (RI-Ei- A-51). . . . . . . . . . . . . . . . . . . . . . . .7. . .

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,. Switchgear Building Construction Rdlk Stoppage....................... 8

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1 PotentitrGenericIssueidTurbine hip Analysis..................... 9

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1 Exit Interview,. .... ....,. 10

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DETAILS

' Summary of Facility Activities The plant operated at full power with the exception of a turbine runback to '

75% power on June 2 The. runback was initiated when Group B Rod 31 dropped into the cora during rod motio The plant returned to full power later that da . Review of Plant Operations The. inspector observed plant operation during regular tours of the following plant areas:

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Control Room --

Security Building-Primary Auxiliary Building

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Fence Line (Protected Area)

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Yard Areas

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Diesel Generator Rooms --

Turbine Building

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Control Point --

Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The-inspector observed various alarm conditions. Operator awareness and response to these conditions were reviewed. Control room and shift manning were compared to regulatory requirements. Posting and control of radiation and high radiation

l areas was inspected. Compliance with Radiation Work Permits and use of ap-propriate personnel' monitoring devices were checked. Plant housekeeping con-trols were observed, including control and storage of flammable material and other potential safety hazards. The inspector also examined the condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencie These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. In addition to normal working hours, the review of plant operations was conducted during midnight shifts, weekends, and holidays on the following day June 20, 1987 9:00 PM to 11:00 PM June 26, 1987 5:00 AM to 7:00 AM July 7, 1987 4:00 AM to ?:00 AM July 11, 1987 5:00 AM to 9:00 AM l

L No unacceptable conditions were identifie Operators were alert and dis-played no signs of fatigue or inattention to dut J

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2 Plant Operations Review Committee (PORC)

The inspector attended several Plant Operations Review Committee (PORC) meet-ings. Technical Specification (TS) 6.5 requirements for required member at-tendance were verified. The meeting agenda included procedural changes, pro-posed changes to the TS and field changes to design change packages. The meeting was characterized by frank dise r 'ons and questioning of the proposed changes. In particular, consideration was given to assure clarity and con-sistency among procedures. Items for which adequate review time was not available were postponed to allow committee members time to review and commen Dissenting opinions were encourage No deficiencies were identifie . Observation of Maintenance and Surveillance Testing The inspector observed various maintenance and problem investigation activi-ties for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment and use of jumpers, per-sonnel qualifications, radiological controls, fire protection, retest, and deportability. Also, the inspector witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instruments-tion was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personnel, procedure details were adequate, and test results. satisfied acceptance criteria or were properly dispositione The following activities were reviewed:

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PMP 9.5-43, Emergency Diesel Generator Cooling Water H'at Exchangers No deficiencies were identifie . Followup on Previous Inspection Findings 5.1 Ultrasonic Examination Results Interpretation (Closed) Unresolved Item (213/84-21-01): Licensee to establish respons-ibility for ultrasonic examination results interpretation. Controlled Routing (CR) 84-1436 was issued to track this item, which initiated a revision of NU-UT-1, Ultrasonic Examination Procedure General Require-ments, Revision 5. The inspector verified that the ultrasonic examina-tion inspection report now includes documentation requirements for ac-ceptance or rejection for each examination. This item is close . 2 Review of Emergency Diesel Generator Jacking Practices

(Closed) Unresolved Item (213/86-29-02): Licensee evaluation and NRC review of the practice of jacking one Emergency Diesel Generator (EDG)

while the other is out of service. The licensee has evaluated this practice and determined that an EDG should not be taken out of service ;

to be jacked while the other EDG is also out of servic This decision I

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has resulted in a change to SUR 5.1-16, Check for Inleakages. The in-spector reviewed the procedure changes and verified that the procedure now specifies in the' prerequisite section that one diesel must be oper-able for the second diesel to be taken out of service to check for water inleakage to the engine cylinders. This item is close .3 Emergency Plan Implementing Procedure Revision (Closed) Unresolved Item (87-03-01): Emergency Plan Implementing Proce-dure (EPIP) Revisions. The inspector reviewed EPIP 1.5-4, Revision 9, 1.5-5, Revision 9, and, 1.5-6, Revision 9 to verify that changes were madt to address NRC concern This review verified that the terminology in the EPIPs was corrected to refer to the NRC's emergency response team leader as the Site Team Leader / Director of Site Operations (STL/DS0).

The EPIPs were also changed to assure that the Director of Site Emergency Operations in the licensee's emergency response organization will inter-face with the NRC STL/DSO prior to issuing event classification messages to the state. This item is close j 5.4 Engineering Expertise on Shift (0 pen) Unresolved Item (85-13-04): The licensee's shift technical advisor (STA) program does not conform to the NRC Commission Policy on Engineer-ing Expertise on Shif The licensee has implemented the dual role shift supervisor (SS)/STA position, as docketed in correspondence to NRC. The licensee's SS/STA's do not hold college degrees. NRC Generic Letter 86-04 requested the licensee to submit plans for meeting the Commission Policy. By letter dated April 22, 1987, the NRC responded to the licen-see's position that, after completion of the current SS/STA qualification program, the individual has the equivalent of a degree. The NRC stated that the Commission believes that dual role SS/STA's should possess de-grees. The licensee is considering alternatives to achieve conformance i

I with tne Commission Policy and has specified a milestone date of August 1987 to accomplish this. This item will remain open pending formaliza-tion of the new SS/STA progra . 5 Surveillance Procedure for Masonry Block Walls (Closed) Unresolved Item (86-25-01): Licensee to develop a surveillance procedure for routine inspection of masonry walls. The need for an in-spection of masonry walls for cracking and general degradation was iden-I tified in IE Bulletin 80-11, Masonry Wall Design. The licensee has implemented Administrative Control Procedure (ACP) 1.0-54, Masonry Block Wall Inspection Program, which delineates responsibilities and identifies required inspection frequencies for safety-related masonry walls. The ACP is designed for use in conjunction with SP-CE-302, Specification for Inspection, Evaluation, and Repair of Cracks in Safety-Related Concrete Masonry Walls, which describes the inspection, documentation, and repair

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processe The inspector reviewed these procedures and found them to adequately provide for safety-releted, block wall' inspection and repai This item is close .6 Specification Procedure for Repair of Masonry Block Walls (Closed) Unresolved. Item (86-25-02): Licensee to implement a procedure governing repair activities on masonry block wall. The licensee has implemented SP-CE-302, Specification for Inspection, Evaluation, and Repair of Crackc in Safety-Related Concrete Masonry Walls. The inspector reviewed this procedure with relation to both this item and:UNR 86-25-01 and found it to adequately fulfill the objectives of IE Bulletin 80-1 This item is close . Followup on Events Occurring During the Inspection 6.1 Licensee Event ~ Reports-(LERs)

The following LERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined whether further information was required and whether there were generic implications. The inspector also verified that the reportir,g require-ments of 10 CFR 50.73 and Station Administrative and Operating Procedures had been met, that appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit Ground Water Inleakage into Containment Cable Vault Due to Inadequate Design

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87-07-01 - Ground Water Inleakage into Cable Vault, Revision 1 The licensee submitted LER 87-07 as.a special report in resp'onse to the identification of a possible substantial safety hazard (PSSH) involving water inleakaoe in the cable vault. This inleakage occurs during.the spring months when the water table in the area is abnormally high. The water.has normally entered through minor cracks in the floor slabs. In

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response to this, the licensee sealed the floo Recently, water has begun entering through gaps between concrete and conduits which are approximately six to eight feet above the floor. Inleakage collects on the floor at a rate of about one eighth inch per hour and could poten-tially submerge five safety related electrical penetrations in the con-tainment wal This was determined to be a PSSH based on the possibility that, during a loss of offsite power, personnel would be distracted with other emergency actions and emergency power would be necessary to remove large amounts of water. Immediate corrective actions consisted of in-structing operations personnel to use the designated Appendix R power sources and pumps for emergency water removal. For long term corrections, the licensee is evaluating several possible methods of dewatering the area around the cable vault buildin .I

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In thehazar safety original LER, the licensee stated that this issue was a substant i If this were an actual SSH, the licensee would be re-quired to submit a 10 CFR Part 21 report per Nuclear Engineering and Operations Part 21. In fact Procedure 2.01, Implementation of the Requirements of 10 To r6flect the nec,essary changes, LERa positive SSH determination has 87-07-01 was issue tion of the PSSH evaluation will be followed during future inspectionsTh .

6.2 _ Employee Found With Marijuana During a routine tour on June 20 at about 2:30 AM, the security shift of the maintenance building in the south yard of the The individual's badge and firearm were removed and he was searche A half The srked guart marijuana cigarette was found in the guard's possessio cas then escorted offsite and his site access was termina In response to this incident, the licensee notified the Connecticut State Polic This 1986. security guard had been hired by the security contractor in hovemb October He had 198 passed his initial physical exam and drug screening in Additional drug testing is routinely conducted on an annual licensee's basis, drugin conjunction testing program.with the annual physicals, as a part of the In response to this incident the lic-ensee has conducted unannounced drug testing of the entire secu,rity force At_the three tested close of this positive forinspection drug usage. period, of 120 security personnel teste those three individual Site access was terminated for Primary Water Storage Tank (PWST) Level Transmitter Line Broken On June 20, radwaste technicians were moving waste storage boxes with a

areafork lift in the south east corner of the radiologically controlled (RCA).

sheared it from the tank.The fork lift hit the PWST level transm A hole was created in the tan The tank water a period drained to the of about ten storm hours. drains and, subsequently, to the river over routine makeup to plant systems.The PWST supplies demineralized water for ,

quires operability of the PWST with 80,000 gallons ofIfwate Technical!

the tank becomes inoperable for more than four hours, a backup source is to be provided or the plant is to be placed in hot standby. According to con-trol room about logs, the Recycle Primary Water Storage Tank (RPWST) contained 135,000 was an adequate gallons backupof water at the time of this event and therefore tan Also, minimum water volume requirements for the Appendix R Safe Shutdown Fire ensure adequate cooling water to This minimum total volume required isshutdown the plant in the e 180,000 gallons of water between the PWST, RPWST, and Demineralized Water Storage Tank (DWST). At the

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i time of this event the DWST contained approximately 93,000 gallons of wate This volume added to the RPWST volume exceeds the minimum re-quired total water volum The inspectors reviewed the discharge permit which was completed after the release. A sample of the water taken during the release indicated that the water contained tritium at a concentration of 4.35 E-06 micro Ci/ml. The licensee stated that this concentration was indicative of the background levels for tritium in the site area. This value is much less than the release limit of 3.0 E-03 micro,Ci/ml set by 10 CFR 20 Appendix The inspector noted that the tritium levels were so low that the discharge could be considered essentially nonradioactive. The in-spector noted further that licensee surveys within the tank on June 21, 1987 prior to repairing the broken level tap identified alpha contamina-tion on one of 11 smears inside the tank. The smear of sediment on the tank bcttom showed 40 disintegrations per minute based on counts recorded on a Ludlum LSC-1 scaler with a Type 43-2 probe. Subsequent chemical analysis of the sediment showed that the alpha contamination was from the naturally occurring isotopes of Ra-226 and Th-22 No unacceptable conditions were identified in review of this event and the licensee's response. NRC discussions of this item with the licensee addressed the undesirability of discharging water via the storm drains and the desirability of notifying the resident inspector for unplanned releases. The licensee agreed that a courtesy report to the resident inspector is appropriate in cases such as thi .4 Dropped Control Rod and Plant Runback At 1:18 PM on June 25, the plant experienced a dropped control rod and the associated turbine runback to 70% power. Operators were inserting rods for a 30 MW power reduction for repairs to the No. 2 feedwater regu-lating valve, when Group B Rod 31 dropped into the core. The turbine runback was initiated by the rod bottom bistable. Load was reduced to 70%. All systems functioned as designed. The cause of the dropped rod was determined to be a blown fuse. The fuse was replaced, the rod was recovered, and the plant was returned to full power by 11:30 PM on June 2 .5 Inoperable Control Rod Position Indication (RPI)

On July 14, at about 2:30 PM, the control room received a rod out of step alarm and indication on the data logger that Group D Rod 41 was 117 steps below its associated bank, which was at 325 steps. Operations personnel verified that there was no apparent core flux change, no turbine load runback, and no core temperature fluctuatfans. SUR 5.3-40, Core Quadrant Power Tilt Determination, was performed. The resulting maximum power tilt ratio was 1.0128; the Technical Specifications (TS) limit is a maximum of 1.0 From these observations the licensee determined that the rod was not mispositioned. The RPI was then declared inoperabl In this situation TS 3.10.F requires that the rod group not be moved

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. 7 greater than eight steps in either direction and that, if position indi-cation is not returned to service within eight hours, the rod must be declared inoperable and the appropriate TS action statement me Troubleshooting commenced immediately. A capacitor in the RPI power cabinet was identified as the cause of this false indicatio Span adjustraents were made and the indication returned to norma As an added verification that the rod was indeed in the correct position, the licen-see conducted a partial core flux map in the position adjacent to rod ,

4 The flux map verified.that the rod was aligned with Group D. The RPI power cabinets are scheduled to be rebuilt during the upcoming re-fueling outage. The inspector reviewed licensee actions in this event and verified compliance with the applicable T No deficiencies were identifie . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported information was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and per-formance specifications; and that planned corrective actions were adequate for resolution of the problem. The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The

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following periodic reports were reviewed:

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Monthly Operating Report 87-05, plant operations for the period May 1-31, 198 Information Letter on Demineralized Water Discharge, dated Jura 30, 1987 No unacceptable conditions were identifie . Cable Connector Qualification (RI-87-A-51)

} During the inspection period a two part allegation was received concerning the adequacy of the environmental qualification for cable connectors in the Core Exit Thermocouple (CET) and Reactor Vessel Level Monitoring (P,VLM) sys-tem Specifically, the allegation stated that, although represented as environmentally qualified by the NSSS vendor, Litton/VEAM cable connectors used in the CET system were susceptible to moisture intrusion. The effect of moisture intrusion in a connector would be to increase the associated CET reading uncertainty. The allegation further stated that similar connectors used in the RVLM system, although retrofitted to alleviate this problem, were still potentially susceptible to moisture intrusion due to gasket and/or con-nector mating surface dimensional inaccuracies. Both of these connectors are used in the Haddam Neck CET and RVLM system i

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The inspector reviewed the technical basis, applicability, and safety signi-fica.nce of the allegation. Both systems provide post-accident supporting information to control room operators. Neither system inputs to a safety system, nor are they relied upon for automatic safety significant actio The CETs input to the Reactor Coolant System (RCS) subcooled margin monitor which is required by Technical Specification (TS) 3.9 to be operable during modes 1,2, and The.. action statement for an inoperable subcooled margin monitor is to hand-calculate subcooling margin every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This informa-tion is used in the event of an accident and is not normally used at powe The RVLM system is used to determine reactor vessel water level during acci-dent conditions. Although this system was installed in response to TMI con-cerns, its operability has not been incorporated into TSs. RVLM inoperability  ;

in the post-accident mode does represent a loss of important informatio However, examination of temperature and pressure data provides another means of determining whether core voiding may have occurred, and radiation measure-ments provide another means of assessing fuel integrit Overall, the CET and RVLM systems, while important, are not essential to pub- i lic health and safety and do not affect the equipment capability or perform-ance levels required to assure safe operation. Also, operations personnel have been informed of the potential for these systems to become inoperable in post-accident condition The alleger stated to the inspector that he had informed the licensee of his concerns. The inspector found that licensee follow-up of the concerns has been initiated. It was also determined that the connectors in question are identified as qualified in the licensee's environmental qualification program for electrical equipmen Because the technical issues raised in the allegation have generic implica-tions, further follow-up of this matter by other divisions in the NRC is anticipate The resident inspector will follow the licensee response to this i

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9. Switchgear Building Construction Work Stoppage On June 17, a special Plant Operations Review Committee (PORC) meeting was conducte Bechtel presented their determination of the cause of the quality assurance problems which resulted in a stop work order last month (see NRC Inspection Report 50-213/87-12, detail 9). Bechtel also presented their pro-posed corrective actions to prevent recurrenc The primary conclusion was that it is necessary to establish a clearer understanding of the safety sig-nificance of this construction and to ensure that the quality assurance pro-cedures are followed. This is to be done by more extensive training and more frequent auditing of the construction process by both contractor and licensee l Quality Assurance groups. After evaluation of this presentation, the PORC agreed to lift the construction stop work order pending the creation of a

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. 9 punchlist of items to be completed before physical construction will begi Items on this list include worker training, design of a dewatering system, and a revised safety evaluation for the excavation and dewatering processe During the week of July 1, the licensee re-evaluated the design and concluded that it was prudent to move the building rather than continue with the plans to dewater the area. This decision was made because of uncertainties involved with the seismic integrity of the Primary Auxiliary Building wall should there be a seismic event during the construction period. The distance of 11.5 feet north has been proposed for building movement so that the required clearance of 20 feet to tne existing security fence is maintained. With this change, there will be many revisions to the design in the PDCRs associated with this building. The inspectors will continue to monitor licensee activities in this are . Potential Generic Issue in Turbine Trip Analysis On June 24, the licensee received an infornal notification from the nuclear steam supply system vendor (Westinghouse) that identified a potentially generic issue with the Final Safety Analysis Report (FSAR) Loss of Electrical Load / Turbine Trip Analysis. This information may require review for report-ability per 10 CFR 50.5 The Westinghouse analysis of the turbine trip operational transient used generic, conservative FSAR assumptions. Analysis showed that the departure from nucleate boiling (DNB) design basis may not be met if a reactor trip does not immediately follow a turbine mechanical trip and a complete loss of forced coolant flow also occurs due to a failure of the fast bus transfer to offsite power. The Westinghouse analysis assumed a standard electrical protection scheme that would result in a fast bus transfer at about 30 seconds after the turbine trip. Even though safety grade reactor trips (e.g., low steam genera-tor level or high reactor pressure) would most likely trip the reactor, the trips may occur just prior to the transfe The resulting postulated loss of flow event would occur with the reactor at or near full power with the reactor vessel inlet temperatures 15 to 25 above nominal conditions. The FSAR complete loss of flow analysis would not bound this event with respect i

to the minimum DNB ratio.

l Westinghouse reported that studies have shown that, while the DNBR design L basis may not be met, no fuel failures due to DNB are expected. Additionally, the Westinghouse evaluation showed that the ONB design basis could be met on a best estimate basis when credit is taken for the reliable, non-Class 1E reactor trip on turbine trip.

l The licensee reviewed the vendor information, determined it to be generally applicable to Haddam Neck, and is evaluating the issue further. Plant In-formation Report 87-70 was written on June 24 to disposition the issue in regard to the immediate impact on plant operations. The licensea noted that there does not appear to be an immediate basis for a concern due to the plant specific design that requires several independent failures to occur to repeat

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,. 10 the scenario used in the vendor's analysis. These include failure of the non-Class 1E reactor trip on turbine trip, failure of two safety grade reactor-trips after the turbine trip,'and failure of-the independent and separate fast bus transfer. schemes for the reactor' coolant pump buses. Additionally, for a turbine trip without an immediate reactor trip from the turbine stop. valves or low auto stop oil pressure trip, reactorLtrip signals are expected to occur at seven seconds'faam high reactor pressure and at 26 seconds from high pres-surizer level. Since the design of the electrical scheme will keep the gener-ator connected to the electrical grid for 52 seconds.following a turbine trip (absent an. electrical fault), the reactor would have been shut down for-about o-26 seconds prior to a postulated loss of forced cooling flow due to a postu-lated failure in the fast bus transfer. Thus, the' transient should not be

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more severe than the limiting loss of flow event analyzed in the FSAR. Based .

on.the above, the licensee concluded that there was no basis for a conclusion '

that a_ potential significant safety hazard existe Licensee resolution of the issue will be tracked via' Controlled Routing 87-635 to obtain. formal evaluation by the NUSCO safety analysis group by September 1, 1987. The inspector identified no inadequacies in the licensee's evalu-ation or intended followup actions. This' item will be followed on a subse-quent routine inspection to determine the final disposition of the matter by the license . Exit Interview During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identified.

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