IR 05000293/1985017
| ML20133N134 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/01/1985 |
| From: | Jerrica Johnson, Mcbide M, Mcbride M, Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20133N089 | List: |
| References | |
| 50-293-85-17, IEB-80-25, NUDOCS 8508130385 | |
| Download: ML20133N134 (15) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N /85-17 i Docket N License N DPR-35 Category C Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: J e 1 , 1985 - July 15, 1985 Inspectors: .d
. Johidon, Senior Resident Inspector 8K 1p,fa 'Date eNes
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Mr Mc de, Resident Inspector Date Approved By: d'. , do Trtn , Chief, Reactor Projects Section 3A 8!![8[
' Da~te Inspection Summary: Inspection on June 13 - July 15, 1985 (Report No. 50-293/85-17)
Areas Inspected: Routine unannounced safety inspection of plant operations in-cluding: Followup of previous inspection findings and NRC Bulletins, operational reports, ESF walkdowns, surveillance and maintenance activities, and health physics activities. The inspection involved 98 inspection-hours by two resident inspector Results: Two violations were identified (Failure to establish and implement sur-veillance procedures, detail 7; and Failure to follow a procedure for radiation work permits, detail 8). One deviation from a licensee commitment concerning the Inservice Test (IST) Program was also identified (detail 6). No other significant safety issues were identifie i
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i 8508130385 050006 PDR ADOCK 05000293 l
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A TABLE OF CONTENTS Page 1 Persons Contacted ................................................... 1 Plant Status ........................................................ 1 Followup on NRC Bulletins and Previous Inspection Findings .......... 1 Operational Safety Verification ..................................... 2 Scope and Acceptance Criteria .................................. 2 Findings ....................................................... 3 ESF Walkdown ........................................................ 6 Scope .......................................................... 6 Findings ....................................................... 6 Followup on Events and Nonroutine Reports ........................... 7 Events ......................................................... 7 Review of Licensee Event Reports (LERs) ........................ 7 Surveillance Testing ................................................ 8 Scope .......................................................... 8 Findings ....................................................... 8 Maintenance and Modification Activities ............................. 10 Scope .......................................................... 10 Findings ....................................................... 11 Health Physics Activities ........................................... 12 10. Management Meetings ................................................. 12
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DETAILS Persons Contacted Within the report period, interviews and discussions were conducted with mem-bers of the licensee and contractor staff and management to obtain information pertinent to the inspectio . Plant Status The plant operated at full power during most of the inspection period. A reactor scram from low power occurred on June 14, 1985 from a high water level indication during low power maneuvers. Power had to be reduced below fifty percent for several days at the end of June due to heavy fouling of cooling water intake screens during stormy weathe . Followup on NRC Bulletins and Previous Inspection Findings (Closed) IE Bulletin (80-25). Operating Problems with Target Rock Safety-Relief Valves (SRV) at BWR This Bulletin was issued to inform all BWR licensees about problems experienced with the two-stage SRVs at Pilgrim in 1980 and to request action regarding 1) inspection of solenoid actuators and testing following an aging period of about 3 months at operating temperatures, 2) revision of procedures to require disassembly, inspection, and testing with steam in the event of failure without clear identification of the cause, and-3) review of the pneumatic supply for the potential for overpressurization, and 4) placement of relief valves and annunciator The licensee provided their initial response to this Bulletin in a letter dated March 19, 198 The following LERs have'also been issued by the licen-see describing these events and corrective actions: LERs 80-30, 80-47, 80-69, 80-79, 80-80 and 81-62. NRC Reports documenting event followup include the following: Nos. 80-25, 80-26, 80-29, 80-30, 81-24, 81-35, 82-10 and 82-1 In addition, a recent NRC Inspection, No. 84-39, documents verification of licensee's actions to prevent overpressurization of the nitrogen supply sys-tem as described in the licensee's letter dated December 4,198 The inspector also verified that the current station maintenance procedure (3.M.4-6 Rev. 12) continues to contain the requirements for disassembly, in-spection, and testing with steam in the event of failure without clear iden-tification of the cause. The Pilgrim station technical specifications were also revised (Sections 3.6.0 and 4.6.0) to provide further assurances of pro-per operation of the SRV The inspector determined that the concerns raised by this Bulletin have been adequately addresse This Bulletin is closed.
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(Closed) Follow Item (84-26-02). Followup on licensee action to resolve dis-crepancies in vessel water level needed to maintain 2/3 core coverage. The licensee submitted a proposed technical specification change, dated June 18, 1985, to correct the 2/3 core water level setpoint listed in T.S. Table 3. The setpoint prevents the diversion of residual heat removal (RHR) system water to containment sprays when the water is needed for low pressure coolant injection (LPCI) to maintain 2/3 core coverage. The setpoint was changed from 302 inches above vessel zero to 307 inches above vessel zero. The licensee submittal indicates that the change is an administrative matter and does not l impact on plant safety. The inspector had no further questions at this time, i
This item is closed.
! (Closed) Violation (84-26-04). Failure to properly review and approve QA-
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related procedures. The inspector verified that four procedures identified
in the violation had been subsequently reviewed and approved by the QA Man-age The following additional corrective actions were verified:
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The licensee completed an initial review of QA program-related procedures by January 29, 1985. The review identified additional procedures which needed QA approval. The approval of the additional procedures was tracked by the QA Departmen The licensee index for QA program-related procedures was updated by March l 1,1985 to incorporate the results of the January review.
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A requirement to annually review issued procedures against the index was
! incorporated into QA procedure 5.0 Station procedure no. 1.3.4, " Procedures", was revised to require that
! QA program-related procedures listed in the index be sent to the QA
, Manager for review and approval.
l The inspector had no further questions concerning QA approval of procedures at this tim This item is close .
l (Closed) Unresolved Item (85-16-02). Review implementation of the Inservice Testing Progra This item is closed for administrative purposes and will
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be tracked under the item of deviation identified in Detail 7 of this repor . Operational Safety Verification Scope and Acceptance Criteria The inspector observed control room operations, reviewed selected logs and records, and held discussions with control room operators. The in-spector reviewed the operability of safety-related and radiation moni-toring systems. Tours of the reactor building, turbine building, intake
- structure, station yard, switchgear rooms, battery rooms, and control l room were conducted.
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Observations included a review of equipment condition, security, house-keeping, radiological controls, and equipment control (tagging).
'b. . Findings (1) On June 18, 1985, the indicator for the "B" Yarway reactor water level indicator in the control room decreased to about +18 inche This instrument shares common indicating and reference lines with safety related level switches for the reactor protection, emergency core cooling, and containment isolation systems. Water level is usually maintained between +24 and +28 inche Other level instru-mentation in the control room did not confirm a decrease in water level. The "B" Yarway indicator returned to normal within a few minutes. The licensee subsequently checked the instrument calibra-tion but could not identify any problems. A signal conditioning board was replaced in the uni On July 14, 1985, the "B" Yarway indicator drifted upwards to +38 inches from an initial value of +28 inches. As before, other con-trol room level instruments did not confirm a decrease in reactor water level and the Yarway returned to normal level within a few minute The licensee stated at the exit interview that personnel located at the local instrument racks in the reactor building did not ob-serve the indicators safety related level switches drifting during the transient on July 14. Personnel did not observe the switches during the first transient on June 18. The inspector had no further questions at this tim The stability of the Yarway reactor water level instruments will be reviewed during future routine inspections of control room activitie (2) On June 24, 1985, the "A" loop of the core spray system was made inoperable in order to upgrade system components for environmental qualificatio The inspector verified that the appropriate proce-dure (No. 8.5.1.4) for one inoperable core spray subsystem was in-itiated and that other emergency cooling system surveillance tests were conducted as required by the technical specifications. The
"A" core spray system was placed back in service on June 29, 198 No inadequacies were identifie (3) On June 24, 1985, the secondary containment system isolated on a false refuel floor exhaust duct radiation signal. The licensee promptly reset the isolatio An investigation indicated that a licensed operator may have inadvertently caused the false radiation signal by jarring the radiation instrument drawer in the control room while conducting a functional test of the radiation monitors.
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! 4 (4) On June 25, 1985, the inspector noted that the "A" logic for group
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I primary containment isolation was tripped. The tripped logic was indicated on a back panel in the control room and the control room operators were not aware of the half isolation at that tim The signal was promptly rese The licensee determined that the isolation was caused by an elec-trical transient in a 480 V a.c. bus (B-17). This bus feeds two 120 V a.c. buses, Y-3 and Y-31, which power a portion of the primary containment isolation system logic. Activation of additional iso-lation logic which was powered by Y-3 and Y-31 was noted at that tim The transient in B-17 occurred when a breaker sparked during post work testing on a core spray valve M0-1400-3A. The licensee evalu-
.ated the condition of B-17, Y-3, and Y-31 after the transient and found no damag All fuses in Y-3 and Y-31 were checked and found acceptabl The inspector had no further questions at this tim (5) On June 29, 1985 at 6:31 p.m., both reactor recirculation loop flows unexpectedly increased while the reactor was at full power. Reactor pressure increased to 1045 psig and power rose to 2034 Mwt (101.8%
of the steady state power limit). The recirculation speed demands for both loops were promptly lowered and power and pressure returned to norma The recirculation loop flows should be independent. While a master controller for both loops is available, this controller has not been used during the current operating cycle because of slight flow stability problems in the "A" loop. Instead, each recirculation loop has been controlled with an independent controller. The lic-ensee evaluated the flow controllers after the transient, but could not identify any problem On July 1,1985 at approximately 11:00 a.m. , flow in both recircu-lation loops, reactor pressure and reactor power unexpectedly in-creased for a second time. Pressure increased from 1032 to 1036 psig and power rose from 1990 to 2012 Mwt. As before, recirculation speed demands were promptly decreased and pressure and power ra-turned to norma No causes for the transients were identified by licensee personnel on duty during the transients. In response to questions from the inspector, the licensee is conducting a more formal evaluation of the incidents. No further transients involving both recirculation loops occurred during the inspection period. The inspector had no further questions at this time. No inadequacies were identified at this time. This is being carried as an Open Item (85-17-05).
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(6) On June 30, 1985 at 6:30 a.m., one of two rupture diaphrams in the steam exhaust line for the high pressure coolant injection (HPCI)
system developed a leak during a routine monthly pump surveillance tes The licensee declared the HPCI system inoperable at 11:55 a.m. on June 30, 1985, following a check of diaphram leak detection instrumentation and a second HPCI test. The licensee notified the NRC of the HPCI problem via the ENS telephone line at 12:45 The HPCI system was declared operable at 5:10 p.m. on June 30, 1985, after the diaphram was replaced and the system successfully ru The licensee stated that the diaphram damage was slight and probably caused by fatigue from repetitive overspeed transients in early June (NRC Special Inspection 50-293/85-16). The diaphram was previously replaced following a HPCI trip and water hammer event in March, 1985 (NRC Inspection 50-293/85-08). The inspector had no further ques-tions at this time. Future HPCI surveillance tests will be reviewed during routine inspections of the licensed progra (7) On July 12,1985 at 3:00 p.m. , a spurious secondary isolation oc-curred during a routine isolation logic surveillance test. The isolation signal was generated when a technician opened the wrong relay contacts at the start of a test of the "B" containment isola-tion logic train, Procedure No. 8.M.2-1.5. The Chief Mainten-ance Engineer indicated that the technician did not follow the test procedure and assumed that the same relay contacts were opened in the "B" logic train test as in the "A" logic train tes The logic test was immediately terminated and the containment iso-lation reset. The NRC was notified of the isolation via the ENS telephone line at 4:58 p.m.. A maintenance supervisor verified that the isolation logic was left in an acceptable state after the ter-minated test on July 12, 198 The technician was counselled on the importance of closely following procedure In addition, all instrument technicians (Nuclear Con-trol Technicians) were informed of the incident and the root caus An inspection of secondary containment dampers was also initiate The Chief Maintenance Engineer indicated that the technician in-volved in the incident was one of the most experienced technicians in the station and had done the surveillance test many times. Sur-veillance test no. 8.M.2-1.5.8.2 was subsequently completed on July 13, 1985. The licnesee plans to submit an LER on the event. The inspector had no further questions at this tim ,
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5. ESF Walkdown Scope On July 8 and 9,1985, the inspector walked down portions of the safety related 125 and 250 V d.c. systems. The electrical lineups were compared to station drawings. Equipment maintenance and surveillance testing were also reviewe Findings (1) An apparent violation of the technical specification involving sur-veillance testing of station batteries is discussed in Section 7 of this repor (2) The backup battery charger for the 125 V d.c. system has been out of service since July,1984. This charger is not required to be operable by the technical specifications. The licensee believes that damaged printed circuit boards in the charger caused it to fail tests in 1984. However, the charger could not be repaired subse-quently using circuit boards that the licensee had in stoc The licensee has received additional boards from the charger manufac-turer and plans to install these in the unit in the near futur At the exit interview, the licensee stated that a safety evaluation was not required to keep the backup charger out of servic The licensee also indicated that differences between the recently re-cuived circuit boards and the older boards in the licensee's stock would be evaluate The inspector had no further questions con-cerning the charger at this tim (3) The inspector noted that breaker no. 21 in a 125 V d.c. distribution panel, 06, was tagged in the open positio This breaker feeds an emergency station lighting panel, 25L. A nuclear Watch Engineer's tag, placed on the breaker in October, 1984, indicated that a QC nonconformance report (NCR) had been issued against a component in the lighting bu Licensee records indicated that an NCR had been issued and cleared during the 1984 plant outage. Quality Control personnel stated that no NCRs were currently open on the 25L pane The licensee promptly removed the tag and energized the breake Personnel from the licensee engineering staff indicated that the emergency lighting on the 25L panel was not used to fulfill the emergency lighting requirements of 10 CFR 50 Appendix R. The lic-ensee stated at the exit interview that QC personnel had not in-formed the control room when the NCR on the 25L bus had been cleare The licensee stated that QC had been reminded to tag control switches as well as local components in the futur .
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The inspector reviewed the startup checklist for the startup from the 1984 outage and verified that the licensee had noted the tag during a system walkdown. The inspector had no further questions at this tim . Followup on Events and Nonroutine Reports Events On June 15, 1985 at 4:35 a.m., the reactor scrammed during maneuvers at about 700 psig reactor pressure. Reactor power had been previously re-duced in preparation for repairs to the turbine control oil syste The scram was caused by an automatic closure of the main steam isolation valves on a high reactor water level signal at greater than 600 psig reactor pressure due to operator erro The licensee conducted a post trip review and' checked the calibration of safety related reactor water level instrumentation. No inadequacies were identified during the review or calibration checks. The reactor was maintained in hot shutdown until later that day when a reactor startup was initiate The inspector observed control room activities while the reactor was shutdown. The high pressure coolant injection system (HPCI) was used in the test mode (recirculate to the condensate storage tank) to draw off steam and cool the reactor. Water was supplied to the reactor via the feedwater system. The HPCI system started smoothly and functioned normally. The inspector also reviewed the level instrument calibration results, pracedure no. 8.M.1-19. No inadequacies were identifie Review of Licensee Event Reports (LERs)
Licensee Event Reports submitted to the Region I office were reviewed to verify that the details were clearly reported and that colrective ac-tions were adequate. The inspector also determined whether generic im-plications were involved and if on site followup was warranted. The following reports were reviewed:
N Subject 85-12 HPCI System Inoperable (LERs 85-12-00 and 85-12-01)
85-13 HPCI Isolation 85-14 Reactor Scram The two HPCI inoperable events described in LER 85-12 were reviewed dur-ing NRC inspections 50-293/85-11 and 85-16. The inspector noted that the cause code for the second HPCI inoperable event was not given in
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either the original.or revised LER 85-12. The licensee agreed to review the second event and submit a revised LER if an additional cause code is neede The HPCI isolation described in LER 85-13 was reviewed during NRC in-spection 50-293/85-11. The reactor scram in LER 85-14 is discussed in Section 6 of this repor The inspector had no further questions concerning LERs at this tim . Surveillance Testing The inspector reviewed the licensee's actions associated with surveil-lance testing in order to verify that the testing was performed in ac-cordance with approved station procedures and the facility Technical Specification A list of items reviewed is included at the end of this report in the attachment to this repor Findings (1) During a review of station battery surveillance test check lists for May, June and July, 1985, the inspector noted one instance where a weekly test of specific gravity, voltage, and temperature had not been conducted for a pilot cell in the station 250 V batter The battery was properly tested on May 26 and June 9,1985, but was not tested during the routine weekly surveillance on June 2,198 The surveillance on the 250 V battery was missed on June 2, 1985 l because the procedural check list, no. 8.C.14, did not require that the battery be teste Instead, the check list required a duplicate test on another battery, the 24 V (A) battery. The licensee stated that the 250 V battery test was inadvertently left off of the checklist when the list was revised in April, 1985. Most of the weekly tests since then had the duplicate 24 V battery test crossed out on the check list and in its place had a 250 V battery test handwritten on the for However, the procedure was not formally change The following instances of failure to follow surveillance Procedure 8.C.14 were also noted:
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Battery temperatures were logged during June 9 and 23, 1985 tests which slightly exceeded the procedural acceptance cri-teria of 77 115 degrees Fahrenheit. An Operations Supervisor and a Watch Engineer had approved the completed tests on both dates. The increase in battery temperatures on these dates did not appear to indicate battery problems, but rather ele-vated ambient room temperature __
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On June 9, 1985, operations personnel may not have realized that the acceptance criteria had been exceeded because a cen-tigrade thermometer was used instead of a Fahrenheit thermo-meter. Procedure 8.C.14 did not contain acceptance criteria expressed in centigrade units. At the exit interview, the licensee stated that Procedure 8.C.14 would be modified to include both Fahrenheit and centigrade acceptance criteri On June 9, 1985, voltage levels for the D17-125 V and D10-250V buses were not entered on the check list as required. On July 9, 1985, 230 V was entered as the 017-125 V battery voltag Battery voltages entries required by the technical specifica-tions were indicated elsewhere on the check lists for both dates (expressed as voltages across the appropriate battery chargers).
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On July 6 and 13, 1985, cells with the lowest specific gravi-ties in the 125 (b) V and 250 V batteries were not chosen pilot cells for testing as required by Procedure 8.C.1 Technical Specifications 4.9.A.2 requires that a pilot cell in the 250 V battery be tested for specific gravity, voltage, and tempera-ture at a frequency of once a week. Failure to establish a proce-dure to implement this test is a violation of Technical Specifica-tion 6.8. Failure to follow the instructions in Procedure 8.C.14 on the dates indicated above is also a violation of Technical Specification 6.8 (85-17-01).
(2) On June 24, 1985, the licensee indicated that one of the pump in-jection valves, M0-2301-8, for the high pressure coolant injection (HPCI) system had not been submitted to the NR The inspector had previously questioned the adequacy of the testing, after noting that the IST program required the stroke time of the valve to be measured in both the open and close directions but the licensee's test pro-cedure only timed the valve in the open directio The inspector requested that the licensee review the IST program to ensure that all tests were implemented. During this review, the licensee noted failures to accomplish the following testing:
Required System Valve Test Frequency HPCI 2301-8 Stroke time to close position Quarterly VRV-9066 Relief valve setpoint verification 5 years VRV-9066 Stroke to open position Quarterly L______________
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Diesel 1 i Stroke to close position Quarterly Oil No. 223 Tran (2 valves)
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Foot-VAL Stroke to close position Quarterly A&B The licensee indicated that the stroke time to close position test for the 2301-8 valve would be added to the routine surveillance procedures. The HPCI relief valve tests will be reviewed by the licensee's engineering department to determine whether the IST tests are feasible. The licensee will also review the diesel oil transfer system to determine if the IST tests are feasibl At the exit interview, the inspector noted the importance of the HPCI relief valve in preventing water hammers in the HPCI steam ex-haust line. Two water hammer events have occurred this year (LERs 85-08 and 85-12). The licensee indicated that a similar vacuum re-lief valve in the reactor core isolation cooling (RCIC) system has also not been teste The RCIC valve is not listed in the IST pro-gram as this system is not safety relate The licensee indicated that both the HPCI and RCIC vacuum relief valves were welded into place and could not be easily removed for testin The licensee may install flange connections for the valves in the future. Setpoint testing was considered during the last out-age, but was not don Neither valve has been tested since the time of original installatio The licensee stated that the original IST program submitted to the NRC, dated April 13, 1979, required the HPCI relief valve set point to be tested every five years. The second 10 year IST program sub-mitted to the NRC, dated July 11, 1983, and a revision to the pro-gram, dated February 14, 1984, also require that the valve be teste Both the original and revised second 10 year IST programs require that the other valves listed in above table be tested as indicate Failure to perform these tests is a deviation from an NRC commitment (85-17-02).
8. Maintenance and Modification Activities Scope The inspector reviewed the licensee's actions associated with maintenance and modification activities in order to verify that they were conducted in accordance with station procedures and the facility Technical Speci-fications. The inspector verified for selected items that the activity was properly authorized and that appropriate radiological controls, equipment tagging, and fire protection were being implemented.
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A list of the items is included at the end of this report in the attach-ment to this repor Findings On July 10, 1985, the inspector discussed recent information concerning the environmental qualification of Limitorque motor operators with the Chief Maintenance Engineer (CME). Tests have indicated that magnesium rotors in some operators may quickly corrode if exposed to a steam en-vironment. The CME stated that General Electric had been asked to de-termine if any of the Limitorque operators installed at Pilgrim have the magnesium rotor The CME indicated that the test data for the magnesium rotors may be misleading and would be evaluated further. The results of the General Electric and licensee evaluations will be reviewed during a future inspection (85-17-03).
9. Health Physics Activities On July 3, 1985, the inspector noted that the frequency of periodic radiation surveillance was not clearly indicated on licensee radiation work permits (RWP) for high radiation areas, as required by station pro-cedures and the technical specifications. For example, RWPs issued for work on the -13 ft level of the radwaste building and for the radwaste trucklock indicated that health physics surveillance was to be periodic, but did not indicate a specific surveillance frequency. Radiation levels in areas controlled by these RWPs were up to 450 mR/h The health physics supervisor who authorized the RWPs indicated that he routinely specified " periodic" surveillance coverage for high radiation areas that did not warrant constant coverag A frequency of " daily when worked" was specified in a " Survey Frequency" box on the RWPs. The supervisor indicated that this frequency meant that health physics tech-nicians had to enter high radiation areas at a minimum frequency of once per shift to provide surveillance coverage and had to conduct a compre-hensive radiation survey once per day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). However, two of four health physics technicians interviewed during the inspection understood the " daily when worked" phrase to mean a minimum radiation surveillance frequency of once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The Chief Radiological Engineer (CRE) indicated that the " Survey Fre-quency" block was only to be used to indicate the frequency of compre-hensive surveys, not the frequency of high radiation surveillance Procedure No. 6.1-022, " Radiation Work Permit", requires that both the survey frequency and the surveillance frequency be entered on RWPs. In addition, the CRE issued written instructions to his staff in February 1985, which required both the survey and surveillance frequencies to be specified on RWP L
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In response-to this finding, the licensee revised all RWPs for high radiation areas changing periodic surveillance coverage to a once per frequency shift. The licensee also discussed the need to indicate a surveillance frequency on RWPs with all health physics supervisors who authorize the RWPs and with the station health physics technician Concerns about the adequacy of RWP control of high radiation area sur-veillance were discussed with the licensee in connection with the first
" chips" incident in 1984 and documented in NRC report 50-293/85-0 Ad-ditional concerns were raised during a subsequent radiation specialist inspection and documented in Report 50-293/84-25. The need to ensure that high radiation surveillance coverage is clearly specified on RWPs was discussed in meetings with licensee management in March 1985 and documented in report 50-293/85-0 Failure to specify a surveillance frequency on RWPs controlling indivi-duals who do not have radiation monitoring instruments or alarming dosi-meters and who enter high radiation areas is a violation of Technical Specification No. 6.13 and licensee Procedure No. 6.1-022 (85-17-04). The following information is included in this report to assist NRC man-agement in following radiation exposure at the station. The monthly personnel radiation exposure for June, 1985 was 66.7 person-rems. The total yearly exposure through June 30, 1985 was 422.1 person-rem . Management Meetings During the inspection, licensee management was periodically notified of the preliminary findings by the resident inspectors. A summary was also provided at the conclusion of the inspection and prior to report issuanc No written material was provided to the licensee during this inspection, other than documents'available in the public document roo .
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e ATTACHMENT TO INSPECTION REPORT 85-17 The following surveillance and maintenance items were reviewed during the inspec-tion perio Portions of the following tests were reviewed:
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Reactor pressure permissive instrument calibration on June 16, 198 J
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Scheduling of ECCS surveillance tests between June 23 and July 1, 1985 while the "A" loop of the core spray system was inoperabl Secondary containment damper inspections on June 23 and July 1, 198 Rod block monitor functional test on June 25, 198 Local power range monitor calibrations on July 2, 198 Weekly tests of station batteries during May, June, and July 198 l
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Quarterly battery test on July 2,198 l
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ECCS and diesel generator surveillance tests on July 7 and 8, 1985 prior to making the "B" loop of core spray inoperabl Portions of the following maintenance and modification activities were reviewed:
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MR 85-24-96, A0N-90 has a cracked drive gea Temporary Modification 85-34, repair drive louver on A0N-9 MR 85-430, HPCI exhaust diaphram high pressure annunciato MR 85-431, replace HPCI exhaust diaphra MR-46-426, low specific gravity L18 Deficiency Tags 2290 and 2333 on 250 V d.c. distribution pane Maintenance on the backup 125 V station battery charger.
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