IR 05000458/1987016

Insp Rept 50-458/87-16 on 870616-0731.Violations Noted. Major Areas Inspected:Licensee Action on Previous Insp findings,10CFR21 Repts,Maint Witnessing,Safety Sys Walkdown, Surveillance Test Witnessing & Inservice Testing of Valves

ML20237K638
Person / Time
Site:	River Bend
Issue date:	08/13/1987
From:	Bennett W, Chamberlain D, Jaudon J, William Jones, Tapia J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20237K590	List: IR 05000458/1987016 ML20237K587 ML20237K611
References
50-458-87-16, NUDOCS 8708190346
Download: ML20237K638 (14)

Text

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APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION I

REGION IV

NRC Inspection Report: 50-458/87-16 Docket.: 50-458 Licensee: Gulf States Utilities Company (GSU)

P. O. Box 220 St. Francisville, Louisiana 70775 Facility Name: River Bend Station (RBS)

Inspection At: River Bend Station, St. Francisville, Louisiana Inspection Conducted: June 16 through July 31, 1987

Inspectors: 3 ' f- h 7 D." Chamberlain, Senior Resident Inspector Date l Project Section A Reactor Projects Branch l l

i N . ow 8' 'l- B'l W.B. Jones,Residen) Inspector Date Proj ect Section A, Reactor Projects Branch 8'-//~F'7 l ,, W. 7:. Bennet t , Project Engineer, Projects Date Section A, Reactor Projects Branch

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etu 6 H - EA- . 'apia Reac' tor spector, Operations Date eipn, actor ty Branch

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Approved: '/J,# '

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.T.Jaujon, Chief,ProjectSectionA Ddte!

Reactor Projects Branch

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, 8708190346 870813 l i l PDR ADOCK 05000458 G PDR L _ __

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Inspection Summary Inspection Conducted June 16 throuah July 31, 1987 (Report 50-458/87-16) j i

Areas Inspected: Routine, unannounced inspection of licensee action on l previous inspection findings,10 CFR Part 21 Reports, maintenance witnessing, safety system walkdown, surveillance test witnessing, operational safety verification, and inservice testing (IST) of valve Results: Within the areas inspected, one viola ion was identified (failure of followup action, paragraph 2).

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i 3 l DETAILS Persons Contacted W. J. Beck, Supervisor, Reactor Engineering

• J. E. Booker, Manager, Oversight

• E. M. Cargill, Supervisor, Radiation Programs J. W. Cook, Lead Environmental Analyst, Nuclear Licensing l

, J. C. Deddens, Senior Vice President, River Bend Nuclear Group

• D. R. Derbonne, Assistant Plant Manager, Maintenance

• R. W. Frayer, Director, Projects

• P. E. Freehill, Outage Manager A. 0. Fredieu, Assistant Supervisor, Operations l

• D. R. Gipson, Director, Quality Services P. D. Graham, Assistant Plant Manager, Operations E. R. Grant, Director, Nuclear Licensing

• J. R. Hamilton, Director, Design Engineering i

• B. E. Hey, Nuclear Engineer, Design Engineering )

G. R. Kimmell, Supervisor, Operations Quality Assurance (QA) R l *R. J. King, Supervisor, Nuclear Licensing

! *I. M. Malik, Supervisor, Quality Systems V. J. Normand, Supervisor, Administrative Services l *W. H. Odell, Manager, Administration

• T. F. Plunkett, Plant Manager

• C. A. Rohrmann, Training Systems Coordinator

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• M. F. Sankovich, Manager, Engineering R. R. Smith, Engineer, Nuclear Licensing

• R. B. Stafford, Director, Operations QA

• K. E. Suhrke, Manager, Project Management R. G. West, Supervisor, Instrumentation and Controls D. W. Williamson, Supervisor, Operations ,

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The NRC also interviewed additional licensee personnel during the inspection perio * Denotes those persons that attended the exit interview conducted on August 3, 1987. Because of schedule conflicts, the Senior Vice President, )

River Bend Nuclear Group, Mr. J. C. Deddens was briefed on August 4,1987, l by the SRI on the results of the inspection. The NRC resident '

inspector (RI), W. B. Jones, and a NRC regional inspector, D. E. Norman, also attended the exit intervie . Licensee Action on Previous Inspection Findings j (Closed) Violation (458/8702-01): Inadequate procedures for directing personnel through the site exit portal monito ,

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The licensee has revised Radiation Protection Procedure (RPP) RPP-0043, i

" Personnel Frisking," and Security Procedure SPI-4, " Security Position 1

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Instructions," to address the required actions by the security officer stationed at the exit turnstile and by radiation protection personnel should the portal monitors alarm. A handout has also been given to all station personnel describing the procedure to be used when exiting through the portal monito i This violation is close (Closed) Violation (458/8509-01): Failure to adequately check calculation ,

1 This item covered four examples of inadequate checking of calculation The first example was a dimensional discrepancy that existed between the computer model for reactor building Support Case R3 and the drawing details. The computer model for Case R3 was corrected and reanalyzed. A review of other computer models was performed to ascertain their accuracy. The second and third examples involved a number of inconsistencies between drawings and calculations for cable tray hold-down connection. The fourth example was a discrepancy in l

the preparation date for a calculation. Examples two through four

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were reevaluated and are not considered a violation. Nevertheless, general retraining of structural engineers was performed to emphasize the importance of the preparer's and reviewer's roles in calculation preparation Based on this retraining and on the reanalysis of discrepant conditions, this matter is considered closed.

(Closed) Violation (458/8509-02): Failure to document and verify assumption This matter involved the qualification of some reactor building cable tray supports using the comparison method of design. The specific l finding was that the process for comparison was not specifically l delineated in detail, and the basis of comparison criteria was i developed, and all reactor building supports not individually designed

were evaluated against the criteria. All supports reviewed were found acceptable. A review of all cable tray calculations showed no other use of unissued reference document Based upon the actions taken in response to this violation, this matter is considered close (Closed) Violation (458/8509-03): Failure to provide basis for I desig ;

The design of cable tray supports outside the reactor building was based on two dimensions (2D) computer analysis results multiplied by ,

factors to make them compatible with the more conservative three !

dimensional (3D) analysis. The basis for the 2D to 3D conversion i factors was not available. This violation occurred because the 1 I

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conversion factors were obtained from an unissued document which did not receive proper review and issuance. In response to this violation, calculations were performed to determine the accuracy of the conversion factors. These calculations were reviewed and found to show that the conversion factors were acceptable. A review of all cable tray calculations was performed and showed no other use of unissued reference document This matter is considered close e .- (Closed) Violation (458/8509-04): Failure to provide adequate disposition of construction deviatio Two corrective action documents were found which provided dispositions based on engineering judgement without sufficient documentation. One disposition involved the attachment of a sloped tray to a cantilever support while the other involved the substitution of internal cross-bracing in lieu of dummy members. These two cases were reanalyzed and the results showed the supports to be adequate. A memorandum was issued to all engineers emphasizing the requirements for documentation of the use of engineering judgemen Based on the response to this violation and the NRC inspectors review, this matter is considered close (0 pen) Violation (458/8531-01): Inadequate integrated leak rate test procedur The licensee's response to this violation dated September 2, 1985, specified corrective action to incorporate changes into surveillance test procedures by December 31, 1986. This specified corrective action was not completed as of June 30, 1987, at which time the licensee subiaitted a revised response to request changing the completion date to December 31, 1987. While the revised date is considered acceptable based on these procedures not being required until the second refueling outage, the failure to complete the original corrective action without prior notification to the NRC was identified as an apparent violation. (458/8716-01)

This item remains ope (0 pen) Deviation (458/8622-02): Failure to install a' solid radioactive waste system as described in the FSA The licensee's response to this violation dated September 19, 1986, specified corrective action which included installation of the sample collection system by the end of 1986 with testing and operation to commence at that time. This sample collection system was still not

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operable as of July 15, 1987. This failure to complete specified corrective action was noted as a second example of the apparent 4 violation (458/8716-01) identified in the previous ite ]

This item remains open The licensee has assembled a QA task force to review all-open commitments ]

made to the NR As a result of the task force effort, a number (12-15) of J overdue commitments (in addition to items "f" and "g" above) were j identified. These items have been evaluated by the licensee and discussed I with the SRI. While no apparent safety concerns resulted from these )

! overdue commitments, the SRI emphasized the importance of meeting )

l commitments made to the NRC. The resolution of the overdue commitments 1 will be evaluated with the licensee's response to the identified violatio . 10 CFR Part 21 Reports The SRI was provided copies of selected 10 CFR Part 21 Reports by NRC Region IV, which may be applicable to equipment or services. supplied to River Bend. These reports were provided to the licensee, who verified that the reports either had been or were being evaluated for applicability at River Ben Any reports that were not already entered into the licensee tracking system were immediately entered. A listing of reports by date,

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manufacturer, and subject is provided below:

February 23, 1987 - GA Technologies, Inc. - Post-LOCA High Range l Radiation Monitor (HRRM) with defective in-containment coaxial cabl I August 13, 1985 - Limitorque Corporations - Worm shaft gear failures on Size 2 Limitorque actuato January 26, 1987 - General Electric (GE) - Agastat GP series relays found improperly seated in their socket February 3, 1987 - Limitorque Corporation - Inadequate instructions for maintenance of torque switche January 23, 1987 - Niagra Mohawk Power Corporation - Inadequate electrical manhole seal design which could allow water into the control building and service water pump bay April 27, 1987 - Static 0-Ring, Inc. - Gas bubble formation between diaphragm layers within sensing element of pressure switches may cause set point to shif ;

The resident inspectors will continue to provide copies of potentially applicable 10 CFR Part 21 Reports for licensee evaluation, and a followup of licensee action on selected 10 CFR Part 21 Reports will be conducted during future NRC inspection No violations or deviations were identified in this area of inspectio . _ _ - - _ _ _ - -

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l 4. Maintenance Witness l

During this inspection period, the NRC resident inspector observed maintenance activities conducted under Maintenance Work Order Requests (MWORs) 103322 and 008745 for work on the electrical protection assembly breakers and the Division III diesel generator lube oil cooler, respectively. In each case, the RI verified through observation and review of records that:

the activities did not violate Limiting Conditions for Operation (LCOs); j the required administrative approvals and tag-outs were obtained '

before initiating work; l

! the procedures used were adequate to control the work; and l the equipment was properly tested before returning to servic I The following observations were made for each of the maintenance activities:

MWOR-103322 - This MWO was initiated to implement Modification Request (MR) 86-0604 which provided for the removal of the reactor protection system (RPS) electrical protection assemblies' overvoltage, undervoltage and underfrequency LED displays. This modification was identified by GE for increasing the reliability of the RPS electrical protection assemblies without affecting plant safety. This MWO was performed during the period of June 19-21, 1987, with the reactor in

operational Condition 4. Prior to the licensee returning the RPS electrical protection assemblies to service, Surveillance Test Procedure STP-508-1700, " Electrical Protection Assembly Semi-Annual Channel Functional Test," was performed and successfully complete I MWOR-008745 - This MWR had been initiated on August 8, 1985, to repair oil leaks on the Division III diesel generator (D/G) lube oil coolers.

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j During the performance of this MWR on June 21, 1987, the RI observed that the mechanics had established and were maintaining the required cleanliness zone around the Division III D/G while the system piping was uncouple The required clearances and work releases were obtained prior to beginning work on the lube oil cooler. A functional test, including inspection of the lube oil cooler, was performed before returning the Division III D/G to servic No violations or deviations were identified in this area.of the inspectio . Safety System Walkdown During this inspection period, the NRC resident inspector performed walkdowns of the "A" and "C" trains of the low pressure coolant injection (LPCI) mode of residual heat removal (RHR), and the low pressure core spray (LPCS) system. These systems are required to be operable during i

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operational Conditions 1, 2, and 3 with an established flow path capable of taking suction from the suppression pool and transferring the water to the reactor vesse The system walkdowns revealed the following:

System valves located on the major flow paths were properly aligned; no abnormal control room instrumentation readings or alarms were present which would prevent LPCI "A," "C," or the LPCS system from responding if required; the RHR and LPCS pump bearing oil reservoirs were properly filled; and the electrical switch gear was properly aligne No violations or deviations were identified in this area of the inspectio . Surveillance Test Witness During this inspection period, the NRC resident inspector observed the performance of Surveillance Test Procedure STP-309-0603, "DIV III 18 MONTH ECCS TEST," on June 23-25, 1987. This surveillance test was performed to verify that the Division III D/G and the associated high pressure core spray (HPCS) system met the following 18-month surveillance requirements:

Functional test of the HPCS system (block valve closed to prevent injection into the reactor vessel);

l proper operation of the Division III D/G following a simulated loss of offsite power in conjunction with an Emergency Core Cooling System (ECCS) actuation test signal; total connected loads to the Division III D/G do not exceed 2600 kw l

and that the auto load sequence timers are operable;

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Division III D/G capable of synchronizing with and transferring its load to the offsite power source; Division III D/G capable of rejecting a load greater than or equal to 1995 kw while maintaining less than or equal to 994 RPM; Division III D/G operating in the test mode, returns to standby operation on an ECCS actuation signal; 4

' Division III D/G operates through a 24-hour run consisting of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 2750 kw and 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at 2500 kw; and Division III D/G is capable of rejecting a load of 2500 to 2600 kw j without tripping and the generator voltage does not exceed 5400 volt i

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l i During the initiation of the HPCS loss of offsite power / loss of coolant- l l accident test, the "C" standby service water pump did not sequence'on i

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within the required 30 (+ or - 3 seconds) following closure of the Division III D/G breaker. The licensee found that the sequencing timer for !

the pump had drifted to a value of 59.8 seconds. The licensee' subsequently verified that the setpoint for the three remaining standby service water pump time delay relays were set within the required tolerance. Following the replacement and calibration of the time delay relay for the "C" standby service water pump, the STP was reinitiated. No further system failures ,

were identified during the performance of the test that resulted in the i HPCS system not meeting the acceptance criteri No violations or deviations were identified in this area of the inspectio ,

i 7. Operational Safety Verification j The resident inspectors observed operational activities throughout the inspection period and closely monitored operational events. Control room activities and conduct were generally observed to be well controlle Corrective actions taken during the previous inspection period to relieve periods of congestion in the control room appear to have been adequat Proper control room staffing was maintained, and access to the control room j operational areas was controlled. Operators were questioned regarding lit l l annunciators, and they understood why the annunciators were lit in all )

l cases. Selected shift turnover meetings were observed, and information j concerning plant status was being covered in each of these' meetings. A walkdown of the "A" and "C" low pressure coolant injections (LPCI) systems and low pressure core spray (LPCS) system was conducted, and the results I are documented in paragraph 5 of this rep' ort. Plant tours were conducted, l and overall plant cleanliness was goo General r;diation protection practices were observed, and no problems were The RI interviewed several individuals inside the radiation control

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note area (RCA) and verified that in each case the individual understood the I requirements of the applicable radiation work permit. Ingress and egress l points for contaminated areas were controlled with contaminated materials l being properly store ')

The SRI and RI interviewed security personnel in the central alarm ;

station (CAS), secondary alarm station (SAS) and in the plant and verified that required compensatory posts had been established when required and that each individual was cognizant of their dutie The resident inspectors also revievied licensee actions on operational events and potential problems. The results of reviews of selected items are described below:

Reactor Feed Pump Minimum Flow Valves: The licensee initiated two condition reports regarding response problems with the "A" and "C" reactor feedpump (RFP) mini-flow valves. The "A" mini-flow valve will not open against condensate system pressure without an operator at the

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\ i valve to assist in opening it, and the "C" mini-flow valve does not close when the associated feed pump is shut dow The licensee has identified these as ongoing problems, which will be corrected during the first refueling outage. Replacement mini-flow valve internals l

have been ordered and are expected to arrive on site by the first week i l of August. The replacement mini-flow valves' pressure equalizing l

valve has been redesigned to correct the problem experienced with the

"A" mini-flow valve. The "C" mini-flow valve has experienced only an intermittent failure to close, for. which the cause has not absolutely {

been determined. This valve will also be replaced with the redesigned ;

mini-flow valve. The "B" RFP mini-flow valve has not experienced failures similar to either the "A" or "C" mini-flow valves. In the i event of a loss of all three feed pumps, the "B" RFP could be  !

restarted from the control room provided the trip conditions have

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cleared, to maintain reactor vessel level without emergency core cooling system initiation. The resident inspectors will monitor licensee actions to correct the mini-flow valves problems during the refueling outag Reactor Scram from High Vessel Water Level: A reactor scram occurred on June 18, 1987, with the unit at approximately 70 percent powe Subsequent investigation by the licensee revealed that the scram occurred because of high vessel water level when feedwater regulating valves locked up on loss of control power. The loss of control power occurred inadvertently during electrical troubleshooting of a nonsafety-related battery inverter which supplies power to balance of plant control and indication circuits. Several contributing factors to the event were identified, but the primary cause was a failure to place the inverter in manual bypass prior to beginning the maintenanc The procedures being used apparently required placing the inverter in bypass, but they did not specify manual or static bypass. The inverter was placed in static bypass which did not prevent the power interruption. Procedures have been revised to require placing the inverter in the manual bypass mode which should prevent a recurrence of this even The SRI reviewed Licensee Event Report (LER)87-012 for this event and an Independent Safety Engineering Group (ISEG) Report 87-005 evaluation of the event. The ISEG report made several recommendations beyond revising the procedures. These included tuch things as additional training on inverter maintenance for operations and maintenance personnel. The NRC resident inspectors will monitor licensee actions to address the ISEG report recommendations during future inspection Main Steam Isolation Valve (MSIV) Closure: On July 4, 1987, with the plant at 76 percent power,=a sing?e MSIV closed during the performance of a surveillance test on the isolation logi Reactor power peaked at 82 percent because of the pressure transient. The surveillance test performance should have caused only one of two solenoids to deenergize and no valve isolation. Subsequent investigation by the

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l licensee revealed that the second solenoid was already deenergized i because of a bad electrical solder joint on the continuity meter to the solenoid. Two_ indications are provided to indicate that the

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I solenoid is energized. These are an indicating light for power l available and an ampere meter for circuit continuity. The surveillance test only required checking of the indicating light on the redundant solenoid; therefore, it was not noticed that the ampere meter was reading downscale. The surveillance test procedures have been changed 1 to require checking both indications prior to test performance. The I defective continuity meter was replaced and the test was subsequently l

completed satisfactorily. No further followup of this event by the

! resident inspectors is planne Division III Battery Discharge: During a weekly surveillance test of the Division III batteries conducted on June 29, 1987, the specific ;

gravities failed to meet Technical Specification (TS) allowable value !

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The batteries were immediately declared inoperable and placed on a.

fast equalizing charge. The batteries were restored to operability on l July 7, 1987. These batteries were apparently discharged during an event that occurred on June 22, 1987, when a transmission line fault caused the Division III electrical bus input breaker to open while the diesel was being test loaded onto the bus. The electrical systems i performed as designed to isolate the diesel generator from the system 1 fault. However, the bus remained out of service during troubleshooting for approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with the battery charger deenergized. The

" pumps and normal loads

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electrical loads on the batteries ',

caused a discharge of the batter <. Th t. % ' of the battery discharge was not recognized c'.i" t aet week, battery surveillance. The licensee hr , t & 'ctiot to pr. tent this type of event from occurring in the fi ;u .: ,, m re. me , procedures were clarified to add strong precal r 1 ding battery discharges and to require immediate specific , ecks if a discharge is suspected. The Division III ba .<ufacturer was also contacted regarding the effects of the disu. . that had occurred. The vendor stated that the batteries were der gni ' to withstand a full cycle or discharge to 1.75 volts per cell nd echarge once per month for the 20-year life. The discharge that .. a occurred was, therefore, not expected to have any detrimental effect on the batteries. No further followup of this event is planned by the resident inspector Reactor Coolant System Leakage Detection: The licensee has experienced problems with diaphragm failures on sample pumps for radiation monitors. A vendor recommendation was made to replace the diaphragms yearly. One of the affected radiation monitors is the diywell radiation monitor which provides two methods of detecting reactor coolant system leakage. The sample pump diaphragm on this monitor had been in service for approximately 2 years. The licensee engineering group had recommended. performing a surveillance on the-pump every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the diaphragm could be replaced. However, TS Section 3.4.3.1 requires at least two methods of leak detection and removing the drywell monitor from service would have left the licensee

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with only one method. of leak detection. This would have required entering TS Action 3.0.3, which requires restoring the monitor. to operable condition within'one hour or placing the unit in the STARTUP mode within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. On July 31, 1987, the licensee requested enforcement discretion from NRC Region IV to allow the plant to remain at power for an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if needed to replace the sample pump diaphragm prior to entering TS Action 3.0.3. Region IV exercised enforcement discretion, and replacement of the pump diaphragm was initiated. The pump repairs were completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with no problems encountere Maintenance Backlog: The SRI continued to monitor overall maintenance  ;

backlog during this inspection period. A significant reduction in the - '

number of maintenance work orders (MW0s) available for work was noted.

l As of July 29,1987, the MW0s available for work was at 800, which is

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below the established goal of 850 and is also at the lowest level since plant licensing. The present level of 800 is down from the recent peak on June 21, 1987, of 1180 MW0s available for work. It is apparent that the additional manpower assigned to maintenance by management has been effective in reducing the overall maintenance backlo No violations or deviations were identified in this area of inspectio . Inservice Testing (IST) of Valves l

l An inspection was conducted to determine whether inservice testing of i valves was being conducted in compliance with the technical and quality l assurance requirements described in 10 CFR 50.55a and Section XI of the ASME 1 Boiler and Pressure Vessel Code, 1980 Edition through Winter 1981 Addend J Gulf States Utilities', " Pump and Valve Inservice Testing Plan-River Bend Station-First Ten Year Inspection Interval-Revision 3," and River Bend Station Procedure No. ENG-3-014, Revision A, " River Bend Standard For ASME XI Inservice Testing of Valves," were reviewed by the NRC inspector to

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verify compliance with the specified test requirements and acceptance

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criteria contained in TS Sections 3/4.6.4, 4.6.4.1, and 4.0.5.

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the test procedures used were the latest ones approved and the test acceptance criteria used was valid for the component being tested; the licensee performs IST in accordance with an approved schedule i within the limitations described in the IST program, including j increased frequency testing; i IST results are recorded and evaluated within the time constraints delineated in the ASME Code; and L-----_------_------_-_-----_____-------------------- - - - .----_--_--_----U

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l IST. procedures and data reflect all requirements of ASME Code l Section XI including:

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evaluation of imposing and removing increased frequency testing i requirements-  ;

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evaluation and justification of changes to' test acceptance criteria, ,

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pump vibration test data analysis and acceptance criteria i justification, including location of vibration measurement;

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requirements that pump tests be conducted at reference I conditions, including reference speed, 1

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compliance of test instruments to 10 CFR 50 Appendix B and ~l ASME Code requirements; i

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performance of positive testing of Category C check valves whose safety function'is to open and close (i.e. , full stroke ,

verification in both directions and individual quantitative leak '

rate testing where applicable);

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evaluation of Category A valve leak test data conducted in accordance with ASME IWV-3426 and 3427 guidelines and including containment isolation and pressure isolation valves; l

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observations of remote position indicators to verify that valve j operation is accurately indicated; and j

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indication that valve streke times are commensurate with the j capabilities of the valve teste i l

The following Station Operating Procedures were also reviewed by the NRC inspector:

i No. STP-204-3307, Revision 4, "Div. I LPCI (RHR) Pump and Valve ,

Operability Test."

No. STP-204-3304, Revision 3, " Loop B RHR Valve Operability Test."

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No. STP-204-0305, Revision 1, " Loop C RHR Valve Operability Test."

No. TSP-0001, Revision 3, " Technical Staff Training and Qualification."

Procedures STP-204-3306 and STP-204-0305 were reviewed in conjunction with the NRC inspector's observation of four valve operability test The valves witnessed are the following:

Valve 1E12*MOVF0378-RHR B to P0OL FPC ASSIST Valve 1E12*MOVF053B-RHR PUMP B SHUT DOWN COOLING INJECTION VALVE l

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Valve 1E12*MOVF0428-RHR PUMP B LPCI INJECTION ISOLATION VALVE Valve 1E12*M0VF042C-RHR PUMP C LPCI' INJECTION ISOLATION VALVE The qualifications and certification to ASME XI-Level II for the test- i engineer in charge were also reviewed by the NR l No violations or deviations were identified during this portion of the-inspectio . Exit and Inspection Interview

An exit-interview was conducted on August 3, 1987, with licensee representatives (identified in paragraph 1). During this interview, the 'l SRI and RI reviewed the scope and findings of the inspectio !

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