IR 05000245/1987011

Insp Rept 50-245/87-11 on 870519-0622.No Violations Noted. Major Areas Inspected:Plant Operations,Surveillance,Maint, Radiation Protection,Physical Security & Fire Protection. Post-outage Spent Fuel Pool Housekeeping Also Discussed

ML20234D757
Person / Time
Site:	Millstone
Issue date:	06/30/1987
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20234D741	List: IR 05000245/1987011 ML20234D737
References
50-245-87-11, IEB-85-003, IEB-85-3, IEIN-84-86, IEIN-87-001, IEIN-87-008, IEIN-87-009, IEIN-87-1, IEIN-87-8, IEIN-87-9, NUDOCS 8707070330
Download: ML20234D757 (15)

Text

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l l U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report: 50-245/87-11 Docket No: 50-245 License No: DPR-21 Licensee: Northeast Nuclear Energy Company Facility: Millstone Nuclear Power Station, Waterford, Connecticut

Inspection at: . Millstone Unit 1 Dates: May 19, 1987 through June 22, 1987 1

Inspectors: Geoffrey E. Grant, Resident Inspector Eben L. Conner, Project Engineer i

Approved: &bd Gl?dl67 E. C. McCabe, Chief, Reactor Projects Section 3B Date Summa ry.: Report No. 50-245/87-11 (May 19 to June 22, 1987)

Areas Inspected: This inspection included routine NRC resident (114 hours0.00132 days <br />0.0317 hours <br />1.884921e-4 weeks <br />4.3377e-5 months <br />), and  !

region-based (22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />) inspection of previously identified items, plant operations, !

surveillance, maintenance, radiation protection, physical security, fire protection, i refueling outage preparations, various IE Information Notices and an allegatio Results: No unacceptable conditions were identified. The licensee's planned post-outage spent fuel pool housekeeping should reduce the potential for recurrence'of undesirable refuel floor radiation level perturbations (Detail 6).

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8707070330 870630 PDR ADDCK 05000245 l G PDR

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TABLE OF CONTENTS

PAGE Persons Contacted.................................................... 1 I 1 Summary of Facility Activities....................................... 1 .

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i Operational Safety Verification...................................... 1 i

' Licensee's Action on Previously Identified Items..................... 2 j.

' IFI 50-245/84-20-01, Reliability of Isolation Condenser Valve 1-IC-3........................................................... 2 ; /84-27-09, Isolation Between Reactor Protection System j (RPS) and Nonsafety Related Equipment........................... 2 IFI 50-245/85-23-04, Incorporation of Blank Badges in Site Dosimetry Quality Assurance Procedures.......................... 3 IFI 50-245/85-24-01, Isolation Condenser High Flow Differential Pressure Switch Snubber 0peration............................... 3 IFI 50-245/85-24-03, Scram Air Header Air Supply Pressure Regulator Performance........................................... 4 IFI 50-245/85-24-02, Site Meteorological Tower Backup Power Supply.......... ............................................... 4 Allegation RI-86-A-122, Main Steam and Feedwater Check Valve Tie-back Rods................................................................. 5 Refueling Outage Activities.......................................... 5 ) Standby Gas Treatment System Actuation.......... .................... 7 Review of IE Information Notice 87-01................................ 7 Review of IE Information Notice 87-08.............................. .. 9 1 Review of IE Information Notice 87-09................................ 9 1 Observation of Surveillance.......................................... 10 1 On-site Plant Operations Review Committee (P0RC)..................... 10 1 Off-site Nuclear Review Board (NRB).................................. 13 l 1 Management Meeting.............. .................................... 13 l

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I DETAILS l Persons Contacted Mr. S. Scace, Station Superintendent Mr. J. Stetz, Unit 1 Superintendent The inspector also contacted other licensee employees including members of the Operations, Radiation Protection, Chemistry, Instrument and Control, Maintenance, Reactor Engineering, and Security Department . Summary of Facility Activities Unit 1 operated at full power during most of the first half of this report period. On June 4, there was a short duration unplanned power reduction to approximately 40% when main conder.ser vacuum decressed due to a failing Steam Jet Air Ejector. Vacuum was restored and the unit was returned to full powe Additionally, routine power reductions were made to accomplish various sur-veillance and maintenance activities. The unit conducted a normal shutdown on June 5 to enter a planned 70 day refueling outage (See Detail 6).

l 3. Operational Safety Verification The inspector observed plant operations during regular and back shift tours of the following areas:

Control Room Cable Vault Reactor Building

Fence Line (Protected Area) l Diesel Generator Room Intake Structure Vital Switchgear Room Gas Turbine Building Turbine Building Control Room instruments were observed for correlation between channels, proper functioning, and conformance with Technical Specifications. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators. Operator awareness and response to these conditions were reviewed. Operators were found cognizant of board and plant conditions. Control room and shift manning were compared with Technical

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Specification requirements. Posting and control of radiation, contaminated 1 and high radiation areas were inspected. .Use of and compliance with Radiation i Work Permits and use of required personnel monitoring devices were checke :

Plant housekeeping controls were observed including control of flammable and other hazardous materials. During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication of equipment status. .

These records included various operating logs, turnover sheets, tagout and jumper logs, process computer printouts and Plant Information Reports. The inspector observed selected actions concerning site security includ ng per-sonnel monitoring, access control, placement of physical barriers, and com-pensatory measures. An inspection of the control room was performed on Satur-L

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day, June 20 from 9:30 a.m. to 12:30 p.m. The operators and shift supervisors' l were alert, attentive and responded appropriately to annunciators and plant conditions. Core off-load was in progress. Correct crew staffing and proper communications were verified. No unacceptable conditions were identifie . Licensee Actions on Previously Identified Items (Closed) IFI 50-245/84-20-01, Reliability of Isolation Condenser Valve l 1-IC-3 This issue relates to licensee review of valve 1-IC-3 design following inoperability of the Crane /Teledyne motor operator due to an improperly I adjusted " closed" limit switch. The valve is oversized for its service, having a 10-inch stroke when only 1-5/8 inches open provides full isola- 1 tion condenser flow. In connection with environmental qualification requirements, the motor-operator for 1-IC-3 (along with seven other Crane /Teledyne motor operators) was replaced with a Limitorque motor-operatcr during the 1985 refueling outage (fall of 1985). The motor-operator selected for 1-IC-3 was a type SMB operator which has a slower speed of actuation than the original operato The inspector reviewed the design package used to replace the eight Crane /Teledyne motor operators, PDCR 1-78-85 (Rev. 1), and the surveil-lance test procedure, SP 623.8. The test data shows the opening stroke time changed from about 2.1 sec to 5.8 sec and the closing stroke time j changed from about 3.2 sec to 6.0 sec after the valve operator was re- '

placed. The Technical Specification (TS) required operating time for 1-IC-3 is less than 10 sec to open (ISI) and less than 19 see to close (containment isolation). The isolation condenser has been automatically initiated without problem since the motor operator was changed. Addi-tionally, periodic routine surveillance of 1-IC-3 has demonstrated con-sistent reliability. No operational problems with 1-IC-3 have occurred since the motor-operator was replaced. The inspector had no further questions. This item is close (Closed) IFI 50-245/84-27-09, Isolation Between Reactor Protection System (RPS) and Nonsafety Related Fquipment NRC Information Notice 84-86 provided licensees with information on potentially significant problems pertaining to the isolation between signals for the RPS and the plant monitoring equipment. This issue is identical to SEP Topic VII-1.A and was addressed in NUREG-0824 (February 1983) and its Supplement 1 (November 1985).

In NUREG-0824, the isolation between the RPS and its power supply was found acceptable as corrected by the licensee. The remaining issues to be resolved were: 1) isolation between the nuclear flux monitoring sys-tems and the process recorders and indicating instruments; and 2) isola-L__ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _

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tion between the average power range monitor (APRM) and the process com-puter. The licensee was to conduct tests to determine if adequate isolation exists for these circuit i By letter dated January 31, 1984, the licensee supplied the results of i their testing and their bases for concluding that modifications are not I warranted. In summary, the licensee concluded that, although isolation )

between the components does not exist, the probability of hot shorts of

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125V dc or 120V ac is remote and electrical failures due to inadequate isolation do not contribute significantly to the overall reactor protec-tion system (RPS) failure probabilit ]

NUREG-0824, Supplement 1, Section 2.15 (SEP) and NUREG-1184, Draft ISAR, Section 3.3.22 (ISAP Topic 1.22) document receipt of NNEC0 letter of January 31, 1984 and state that this information is currently under re-view. The inspector had no questions or issues beyond those being re-viewed by NRR under ISAP Topic 1.22, Electrical Isolation. Therefore, this item is close I c. (Closed) IFI 50-245/85-23-04, Incorporation of Blank Badges in Site ;

Dosinietry Quality Assurance Procedures

1 The licensee routinely sets out extra dosimetry badges in various areas )

of the plant to monitor background conditions. These badges are pro- !

cessed along with personnel and spiked badges by the Dosimetry Processing Laboratory. The extra badges are used to trend background reading At issue was the lack of procedures for this quality assurance measur !

The licensee subsequently developed procedure HP 947/2947/3947, Area Monitoring, to control use of badges for monitoring background condition The inspector reviewed the procedure and determined that it adequately )

addressed the procedural control concern. This item is close l

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d. (Closed) IFI 50-245/85-24-01, Isolation Condenser High Flow Differential i Pressure Switch Snubber Operation l

During testing of the Isolation Condenser, one of the four high flow I differential pressure switches' overshot the full flow value. These de-vices provide protection in the event of a line break in the Reactor Building by isolating the Isolation Condenser from the reactor. The licensee committed to investigating the instrument surge damping snubbers during the next refueling outag Licensee investigation of the surge damping snubbers revealed unequal damping on the high flow differential pressure switch that had previously overshot its full flow value. During the outage, all of the surge damp- 1 ing snubbers on the Isolation Condenser High Flow Differential Pressure Switches were replaced and adjusted to provide equal damping. The Isolation Condenser was placed in service and correct instrument response was observe Additionally, SP-412L, Isolation Condenser Isolation In-

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. 4 I strument Functional and Calibration Test, was modified to include bi-annual testing of the surge damping snubbers. This problem has not re-curred during subsequent use of the Isolation Condenser. The inspector reviewed the licensee's corrective actions and found them to be compre-hensive, timely and effective. This item is close e. (Closed) IFI 50-245/85-24-03, Scram Air Header Air Supply Pressure i Regulator Performance I A reactor trip occurred on October 7, 1985 due to low pressure in the scram air header. The plant is protected against scram valves drifting-open in the event of low air header pressure by four pressure switches

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set at 54 psig (normal header pressure is 70 psig). The scram occurred i during surveillance testing which required tripping individual Reactor i Protection System (RPS) channels and thereby causing one-half scram One or more solenoid-operated scram pilot air valves ("117" and "118"  !

valves) failed to fully reposition when the first one-half scram was  ;

reset. This air leakage path, along with a dip in air pressure caused I by attempting to reset the second RPS channel one-half scram, resulted in a momentary low pressure condition in the air supply header. This condition caused a reactor scra !

The licensee repaired the identified faulty "117" and "118" valves and  ;

resumed power operations. The licensee has subsequently rebuilt all of '

i the "117" and "118" valves to ensure proper operation. The issue re-  :

l maining was investigation of the possible contribution to the scram of '

i malfunctioning air supply header pressure regulating valves and their associated filters. Subsequent licensee investigation determined that performance of the air supply pressure regulators and filters was satis- a factory and had not contributed to the scram. The licensee's investiga- i tion, analysis, immediate corrective actions and long term maintenance

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program effectively addressed this issue. This item is close f. (Closed) IFI 50-245/85-24-02, Site Meteorological Tower Backup Power Supply The site meteorological tower (MET) power supply is a 23KV, above ground line. Power to the tower was lost during a hurricane on September 27, 1985, and the MET tower was out of service for approximately 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> With meteorological instruments out of service, the licensee relied upon back-up data from a contract service at another location in accordance '

with emergency action procedure The licensee committed to reviewing the capabilities of the present system and the need for a backup power suppl Initial licensee review indicated a more reliable power supply would be beneficial. Additionally, because operability of a radio repeater on the tower is required for 10 CFR 50 Appendix R considerations, an alter-nate source of power is needed. Consequently, during the current re-

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fueling outage, a gas generator backup power supply is being installed at the tower. Its operability will be examined incident to routine resident and specialist inspection. The inspector had no further ques-tion . Allegation RI-86-A-122, Main Steam and Feedwater Check Valve Tie-Back Rods liiis October 1986 allegation, regarding modification work by Ebasco in 1980, was that inadequate pipe break restraints existed for main steam and feedwater valves in the Millstone 1 main steam tunnel. This reportedly resulted from the replacement of check valves in the main steam and feedwater lines with physically larger valves. The concern was that tension bar rearrangement, :

which made room for the larger check valves, introduced additional stress due I to an induced bending movement. This issue was transmitted to the licensee by letter dated November 7, 1986. NNEC0's response was issued on December i 15, 198 The inspector reviewed the December 15, 1986 response and discussed the alle-gation with the responsible engineer. The allegation that modifications were l required to account for larger feedwater check valves was true. Both attach -

ment points on the anchor frames and the pad eyes welded to the flued heads j were moved to provide correct alignment. These changes were well documented arid photographs of the modifications were available for review. In an October ]

l 16, 1986 follow-up call, the alleger stated that the problem with the feed-l water line supports had been corrected in 1980, but that the steam check valve ;

l sJpports had not been correcte This inspection confirmed that the only main steam valves in the main steam

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l tunnel are the four outboard main steam line isolation valves (MSLIVs). There {

are no steam line check valves and no modification work was done on the MSLIVs i in 1980. Therefore, the allegation that check valves of a larger size were ;

installed and improperly supported in the steam lines is unsubstantiate The inspector concluded that no safety inadequacy existed in this cas . Refueling Outage Activities On June 5, Unit 1 commenced a 70 day refueling outage with a scheduled com-pletion date of August 14. Major outage activities include: i

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Jet r Instrumentation Nozzle Replacement - the existing assemblies have en degraded by Inter-Granular Stress Corrosion Cracking (IGSCC)-

and will be replaced by new penetration assemblies that are resistant I to IGSCC attack.

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Process Computer Replacement - also implements Safety Parameter Display

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System (SPDS) and a new microprocessor-based Rod Worth Minimizer (RWM).

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Refueling - includes full core off-load, replacement of 196 fuel assem-blies with GE8B fuel, replacement of four control rods, replacement of ,

failed /end-of-life Local Power Range Monitors (LPRMs) with GE NA300 '

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Inservice Inspection (ISI) - includes ultrasonic (UT) inspection of 160 .

stainless steel piping welds in the recirculation, shutdown cooling, J reactor water clean-up and isolation condenser systems that are suscept-ible to IGSCC. Additionally, 45 welds in the non-QA, non-safety related 1 portion of the RWCU system will be UT inspected. ASME required UT, ]

penetrant testing (PT), and magnetic testing (MT) inspections of 115 1 items are scheduled. Automated UT inspections of ten reactor vessel 1 nozzle-to-vessel welds and nozzle inner radius inspections are schedule j Remote visual inspection of the CS spargers, IRM/SRM dry tubes, moisture separator and reactor vessel will be accomplished. Shroud hold-down l j

bolts will be remotely checked (UT) per GE SIL 433. Feed and condensate !

piping will be tested for thicknes UT of over 1,000 piping hangers will be performed to fulfill a 1985 outage commitment. UT inspection l l' I of the drywell shell will be performed in response to Generic Letter 1 87-0 MOV Replacement - the motor operators on RR-2A and 2B and IC-2 will be replaced with Limitorque operators to fulfill EEQ requirement ,

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10 CFR 50 Appendix R Modifications - includes Unit 1/ Unit.2 backfeed, '

control room Halon system, alternate shutdown cooling, emergency lighting, I i hydrogen system piping, reactor pressure and level instrumentation, radio i l

repeater back-up power, and Control Rod Drive (CRD) pump l

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Main Turbine Rotor Replacement - replaces the "B" Low Pressure rotor (which currently has some wheel cracking) with an improved monoblock design roto Overcurrent Trip Device Replacement - replaces electromechanical trip devices on thirty safety-related 480V breakers with Micro Versa solid state trip device Torus to Drywell Pumpback System.- modification to allow the drywell nitrogen compressor to take a suction on the torus and discharge to dry-well. This will reduce the need to frequently vent and purge to maintain the required 1 psid between the torus and drywel Standby Liquia Control System Upgrade - implements use of Boron-10 en-riched sodium pentaborate. This is to fulfill requirements of.10 CFR 50.62 (ATWS).

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Chemical Decontamination process will decontaminate the recirculation and RWCU systems with an expected savings of approximately 400 man-rem.

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Containment Integrated Leak Rate Tes IEB 85-03 Response-M0 VATS testing of safety-related motor operated valve Insulation Replacement - replaces cracked NORYL insulation on safety-related bus wor _ _ _ - _ _ _ .

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Seismic Hangers modifications to 160 seismic hangers on various Cate-gory I System Ventilation Modifications - connects feedwater coolant injection (FWCI)

and condensate pump area ventilation coolers to a source of emergency ?

powe i 7. Standby Gas Treatment System Actuation On May 27, the Standby Gas Treatment (SGT) System automatically actuated on i refueling floor high radiation level (>100 mR/hr). The high radiation level '

was a momentary spike and quickly returned to a normal level of approximately 10 mR/hr. After area and airborne radiation surveys of the refuel floor, ventilation system and plant stack showed normal levels, the SGT system was returned to normal standby service. The high radiation level coincided with placing the spent-fuel pool cooling system in service following valve main-tenanc Refilling the system forced air into the spent fuel pool via the sparge The bubbles rose in the pool, causing agitation. The licensee !

postulated that the agitation caused highly contaminated vacuum cleaner filter I cartridges, which are stored in the pool, to be forced up near the surfac These filters contain debris from past reactor vessel cleaning evolutions and ,

are cuspended by nylon line in the spent fuel pool. With the filters near i the surface, water shielding would have effectively been decreased, causing increased radiation levels. Dosimetry of personnel in the pool area at the ] '

time of occurrence showed exposure consistent with normal general area radi- ,

ation level I I

The licensee is reviewing spent fuel pool cooling system operation procedures I to determine if improved methods of refilling the system are possible. Addi- !

tionally, the licensee intends to perform spent fuel pool general housekeeping after the current refuel outage. These activities will remove disposable equipment and ensure that remaining items are suspended by wire cable vice nylon line. During routine inspection activities, the adequacy of these measures to prevent recurrence will be reviewed. The inspector had no further questions at this tim '

8. Review of Information Notice (IN) 87-01 I Information Notice 87-01, RHR Valve Misalignment Causes Degradation of ECCS in pressurized water reactors (PWRs), alerts licensees to a potentially sig-nificant problem pertaining to residual heat removal (RHR) valve alignment in the low pressure emergency core cooling system (ECCS). The problem con-cerned low pressure ECCS valve abnormal alignments to support testing and maintenance. These placed the system in a degraded configuration uutside design bases assumptions. Although addressed to PWR facilities, the issues raised in this Notice have potential generic implications for boiling-water reactors (BWRs).

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The resident inspector reviewed the Unit 1 Low Pressure Coolant Injection (LPCI) and Core Spray (CS) system designs, operating characteristics and operating histories to determine the applicability of IN 87-01 concerns. The separate and redundant CS trains independently supply low pressure accident cooling to the reactor. Either train is sufficient to meet cooling require-ments. Testing of the CS system is performed one train at a time, with restoration completed between tests. Additionally, if a genuine accident signal was received during normal CS system testing, key motor-operated valves would automatically realign to the accident mode configuration, allowing ,

i appropriate system operation. Thus, the CS system is generally not suscept-- !

ible to the problems addressed in the Notic Although the LPCI system is separated into two loops, each containing two pumps, both loops must be operable and the LPCI discharge line cross-tie valve must be open for the system to fulfill its ECCS functio In order to pro-- J vide required design basis accident core cooling flow, a minimum of three of the four LPCI pumps must discharge to the core via-the unaffected recircula-tion system loop piping. (Note: LPCI is redundant to the CS system. Either i train of CS', or the LPCI system by itself, will provide sufficient low pres-sure ECCS flow.) Under accident conditions, the LPCI Break Detection Logic prevents LPCI discharge to the affected recirculation loop and diverts af- J fected side LPCI flow, via the LPCI cross-tie line, to combine with the un-affected side LPCI flow in discharging to the associated recirculation loo Thus, if the cross-tie valve (LP-8A) is closed, two pump injection flow would be the maximum available and would be insufficient to meet LPCI design basis requirement Like the CS system, during normal system testing, key motor operated LPCI valves would automatically realign to the accident moda configuration if an i accident initiation signal was present. This feature does not apply to LP-8A because it does not respond to an initiation signal. Lowever, LP-8A is a-normally open motor-operated valve controlled from the LPCI- panel located in the control room and is keylocked in the open position with the key removable only in the open position. To demonstrate operability, LP-8A is routinely j cycled on a monthly basis during the conduct of SP622.7. LPCI System Oper- !

ability Test. The closing and opening strokes are each required to be less l than 120 seconds. Therefore, under normal concitions, the maximum time that LP-8A is not fully open is roughly four minutes per month. If the valve i

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failed to reopen remotely during surveillance testing, it is accessible for opening by local use of its manual operator. T1e inspec~.or found Unit 1 not to be susceptible to the concerns in IE Notice 87-0 ,

Licensee review of this matter also concluded that the sprific concerns ,

addressed in this Notice were not applicable to Unit 1. Their review of the l Notice's generic applicability concluded that acpquate measures existed to- !

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! Review of IE Information Notice 87-08 i

IE Information Notice 87-08, Degraded Motor Leads in Limitorque DC Motor-Operators, alerted licensees to the existence of potentially defective DC motors installed in some Limitorque motor-operators. The identified motors contain leads that are susceptible to insulation degradation and subsequent short circuit failur Licensee field review of Limitorque motor operators determined that none of the potentially defective motors were installed in safety-related valves, j The type of valve operators that use these motors are not installed in the plant. However, one of these valve operators is scheduled for installation on IC-2 during the upcoming refueling outag The motor on this operator has i been verified by the licensee as not being subject to this concern. The licensee also has two of the subject motors as spares stored under corporate l

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control. These spares could be used to replace the motor on the IC-2 operator i if it should ever fail. These motors have been physically tagged to' indicate j the requirement for repair prior to transfer from corporate to site cognizanc j The inspector had no further questions in this are j 10. Review of IE Information Notice 87-09 IE Information Notice 87-09, Emergency Diesel Generator Room Cooling Design l Deficiency, alerted licensees to a potentially significant problem involving {

degradation of Emergency Diesel Generators (EDGs) following a loss of offsite l power (LOOP). The issue concerned a potential common-mode loss of EDGs due  !

to isolation of EDG room cooling air supply during a LOOP. In the postulated scenario, the EDGs failed due to high ambient temperature resulting from isolation of room cooling air flow. Air flow would be lost when pneumatically i operated duct dampers closed upon loss of air. pressure. Damper operating air was supplied by a non safety air compressor which would be lost on a LOO The resident inspector reviewed the" EDG and Emergency Gas Turbine Generator (EGTG) room cooling air designs at Unit 1 to determine the applicability of this concer The building housing the EGTG has permanently open louvers i providing a free flow of air through the buildin The turbine and loading gear compartments (which closely surround the EGTG) are cooled by outside air i forced through the compartments by ventilation fans powered by the EGTG Motor '

Control Center (MCC). Dampers on the ventilation exhaust are held open by latches during normal operation. The latches release, allowing the dampers -

to close, upon actuation of the Carbon Dioxide Fire Protection System. The '

EDG room is cooled by room air handling units. These units recirculate room air through cooling coils that are. supplied by the Turbine Building Secondary Closed Cooling Water (TBSCCW) syste TBSCCW pumps are powered from 480 volt safety buses. This ventilation unit does not have any dampers. Fans that recirculate air through the units are powered from an MCC powered by the ED i

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Thus, neither emergency power supply is susceptible to a common mode cooling failure on a LOPP. In fact, EDG and EGTG room cooling designs are fairly resistant to simple failure. Licensee review of this matter also concluded that the concerns addressed in the Notice were not applicable to Unit The inspector had no further questions in this are . Observation of Surveillance On two occasions (one during the last report period), the inspector observed parts of the weekly battery surveillance. The Switchyard Battery. surveillance was reviewed in relationship to a proposed TS change being processed by Region I. The change would delete the switchyard batteries from the TS allowing surveillance to be performed by the New London Area Substation Equipment Maintenance Group (NLSEMG), who has the surveillance responsibility for the other switchyard equipmen The inspector reviewed the following procedures, j

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SP 780.1 - Switchyard and Station Battery Weekly Inspection

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SP 780.2 - Switchyard and Station Battery Quarterly Inspection

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SP 780.3 - Load Test on Station Batteries

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SP 780.4 - Station Battery Service Test

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SP 780.5 - Switchyard and Station Battery Annual Inspection

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1-C T6-3 - Stationary Battery Installation & Operating Instructions (NLSEMG Substation Batteries and Chargers Surveillance Procedure )

The SP 780 series and NLSEMG substation procedures are similar in conten The switchyard battery surveillance was performed adequately without direct use of the procedure although, during both observations, the electrician had the procedure with him. No unacceptable conditions were identified. However, che inspector noted two areas where the procedures could be made clearer, j First, the selection criteria for a new pilot cell could be more clea i Second, procedures lack some clarity on the process of conducting battery equalizing charges. These two area: were discussed with the responsible engineer, who is in the process of rev4 sing the proceduras. These matters will be reviewed again incident to routine inspection. The inspector had no  !

further question ;

12, On-Site Plant Operations Review Committee (PORC)

The resident inspector attended Unit 1 PORC meetings on May 20, 22, and 27, and June 4, 5, and 9. Technical Specification 6.5.1 requirements for com-mittee composition were met. PORC topics included the following:

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Discussion of methodology for dewatering /desludging the coisture separa-tor / dryer pit. Several options were considerea. Radiological conse -

quences of various approaches to dewatering were discussed. Excellent interplay among PORC members yielded a well-defined solution to the probler 1

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Special test procedure for new Rod Worth Minimize PORC questioned i operability requirements'during testing and was satisfied with procedural !

coverage of this issu !

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Plant Design Change Request (PDCR) for installation of control room Halon fire suppression system. Extensive discussion included analysis of sys- ,

tem operation, interlocks and safety precautions. PORC members concurred J with the findings of applicable safety evaluations. Active discussion thoroughly covered this PDC A 10 CFR 50 Appendix R consideration was discusse The licensee was submitting a letter to the NRC stating their position relative to use )

of Mineral Insulated (MI) cabling. If the NRC does not concur with this )

position, then the letter will be the basis of an exemption reques !

PORC discussion centered on MI cable applications in the plan j

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Proposed Technical Specification Change Request to section 3.5.C.2 that

increases the minimum water inventory in the Condensate Storage Tank from 225,000 to 250,000 gallons. PORC questioning determined that the request !

was necessary for 10 CFR 50 Appendix R considerations for safe plant !

shutdown after a fire in the Reactor Building. This topic generated 1 active dialogue among PORC members concerning the nature of Appendix R requirements and consideration j

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A new procedure for identifying and processing minor plant changes was discusse The new procedure examines proposed changes / work to determine if they warrant treatment as a PDCR or if a less formal method is appro-priat The procedure requires that the results of this evaluation be presented to PORC to obtain concurrence. Several questions were raised and satisfactorily answe rec'. The new procedure was implemented and several of the required evaluations were presented at PORC meetings. ".his appears to be an effective method for unencumbering the design change process while retaining an appro-priate level of safety oversigh Review of several minor and routine operation, maintenance, surveillance and chemistry procedure change PDCR for provision of a fire water connection to the Reactor Building Closed Cooling Water (RBCCW) system and the Low Pressure Coolant Injec-tion (LPCI) system heat exchanger. These changes were precipitated by 10 CFR 50 Appendix R requirements for alternate cooling availability in various fire scenarios. There was excellent PORC discussion of hose re-quirements, hose storage locations, use of the new capability and post-modification testing requirement _ _ _ _ - _ - _ _ _ _ _

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PDCR for installation of an excess flow check valve in the Hydrogen sup-ply header in the Turbine Building. That valve would limit the maximum !

building hydrogen concentration in the event of a pipe break. PORC questioned the valve location and tabled further action pending resolu-tio Special procedure for chemical decontamination of recirculation system piping. In conjunction with the PDCR on this program, this procedure detailed the decontamination process, conditions and precautions. De-tailed discussions by PORC .wmbers ensured their thorough understanding of this evolutio Special test procedure to determine Jet Pump Instrument (JPI) operability following installation of the new JPI nozzle PORC members assured themselves that the test adequately determined operability.

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PDCR for installation of Torus to Drywell nitrogen pumpback system. This system is designed to reduce the frequency of venting and purging the Torus and Drywell while maintaining the required differential pressur PORC actively pursued issues related to this PDCR including system normal operation, response to containment isolation signals, safety concerns, and alternate design possibilities. PORC concurred with the PDCR safety evaluation PDCR for application of Mineral Insulated (MI) cable in reactor pressure and level indicating circuits to meet 10 CFR 50 Appendix R requirement Extensive PORC discussion covered quality of materials, the necessity for design change, and operability and precise identification of instru-i ments requiring this change. PORC approval was postponed pending better l identification of the particular pressure and level indicators involve PDCR for modification of Control Rod Drive (CRD) pumps to allow self- )

cooling using pump discharge flow. This planned modification provides 4 a reliable source of high pressure make-up water to the reactor in the event of fire and in accordance with Appendix R. PORC tabled approval'

l pending modification of the design to include double isolation valves.

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Review of an LER/PIR concerning incomplete installation of passive fire j coating in the Emergency Diesel Generator roo PORC covered immediate j and long-term corrective action l

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PDCR for replacement of CS-21A & B motor operato A new Limitorque l operator should provide more reliable valve operation at higher differ- i ential pressures. PORC discussion covered the safety evaluatio ]

The inspector observed that PORC e, embers exhibited probing and questioning attitudes. They effectively used discussion periods to focus attention on the safety implications of presented items. Active interplay among members supported the PORC chairman in making meaningful and informed decision ;

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i Special presentations made by various staff representatives was an effective j method of broadening PORC's understanding of several technical issues. The inspector had no further questions in this are . Off-Site Nuclear Review Board (NRB)

The resident inspector attended a Unit 1 NRB meeting on June 9. Technical i Specification 6.5.3 requirements for board composition were met. NRB topics  !

included the following-

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Previous meeting minute Outstanding action item !

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NRB audit schedul j l

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Two proposed Technical Specification Change Request i Discussion of new Administrative Control Procedure that streamlines the  ;

modification process (see detail 12).

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LER 86-028, Revision Two I&E Inspection Reports including discussion of one security related

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1 1 violatio Several Plant Design Change Request (PDCR) safety evaluation ( Presentation by Unit 1 Engineering of the PDCR development process.

The inspector observed that the NRB performed Technical Specification 6.5. t required reviews at a level consistent with the safety significance of the issu Discussions were consistently perceptive and professiona The-in- -

spector had no further questions in this are . Management Meetings

At periodic intervals during this inspection, meetings were held with senior plant management to discuss the finding No proprietary information was l

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identified as being in the inspection coverage. No written material was provided to the licensee by the inspecto P s

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