IR 05000293/1985008

Safety Insp Rept 50-293/85-08 on 850402-0506.No Violation Noted.Major Areas Inspected:Plant Operations,Including Previous Insp Findings,Operational Safety Verification & Surveillance & Maint Activities

ML20129G641
Person / Time
Site:	Pilgrim
Issue date:	05/28/1985
From:	Jerrica Johnson, Mcbride M, Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20129G604	List: IR 05000293/1985008 ML20129G599
References
50-293-85-08, 50-293-85-8, NUDOCS 8506070326
Download: ML20129G641 (13)

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U.S. NUCLEAR REGULATORY COMMISSION DCS Nos. 50293-850331 REGION I 50293-850404

, Report N /85-08 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: A ril 2 1985 - May 6, 1985 Inspectors: ;<. [.fdd

. Johnt>6n, Senior Resident Inspector kB6

' [Iate 1 7 $ Ath$

K. McBr 54, Resident inspector

$' Uate h5 Approved By: b. uk0 L f rippl' Chief, Reactor Projects '

f!8f Date Section No. 3A, Projects Branch No. 3 Inspection Summary: Inspection on April 2, 1985 - May 6, 1985 (Report N /85-08)

Areas Ins 3ected: Routine unannounced safety inspection of plant operations in-cluding: ;ollowup on previous inspection findings, operational safety verification, followup of events and non-routine reports, surveillance and maintenance activi-and a review of licensee actions ties,chemistryandhealthphysicsactivities}nmentsystem regarding the use of liquid nitrogen in conta The inspection in-volved 134.5 inspection-hours by two resident inspector Results: No violations were identified.

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DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with members of the licensee and contractor staff and management to obtain the necessary information pertinent to the subjects being inspecte . Plant Status The plant operated at full power during most of the inspection period. A raactor scram occurred on April 4, 1985 following a trip of the main turbin Power operation was resumed on April 5, 198 . Followup on Previous Inspection Findings (Closed) Violation 293/83-09-01).

foruseduringrece(iptinspectionwithoutmeetingthedocumentationandSaf identification requirements of Purchase Order 62102. NRC Report No. 83-19 documents initial review of this item and the implementation of cor-rective actions described in the licensee's response dated July 6, 198 TheinspectoralsoverifiedthatthegualityAssuranceManual,Section 15.2.5, requires the Nuclear Engineering Department and Quality Assurance Department to review and approve " accept" or " repair" dispositions to Nonconformance Reports (including those written for receipt inspection).

This item is closed, (Closed)FollowItem(293/83-24-02). Check that acceptance criteria are included in the diesel generator fuel oil surveillance procedure. The inspector reviewed procedure no. 7.1.3.6, " Diesel Generators' Fuel Oil Sampling and Quality Analysis", Revision 4, and noted that ASTM accept-ance criteria are included in the procedur The inspector had no fur-ther questions. This item is close (Closed)UnresolvedItem(293/84-24-01). Clarify inspection requirements on work plan and inspection record (WP & IR) documents. The WP & IR procedure no. WP/P-1, was revised following the 84-24 inspection to in-cludeaddItionalinstructionsfordocumentingWP&lRinspectionrequire-ment The inspector reviewed the revised procedure and a WP & IR for installing hangers for electrical conduit, No.84-060.05-E-152. No un-acceptable conditions were identified. Additional WP & IR's will be re-viewed during future routine inspections of the licensed program. This item is close (Closed) Violation (293/84-36-01). Failure to maintain the "B" source range monitor (SRM) operable during fuel movement in that quadrant. The licensee's response to the violation, dated January 9 1985, stated that a " Fuel Load Checklist" would be added to operator shlft turnover sheets to help ensure that information such as bypassed SRM's would be trans-mitted between shifts. The inspector verified that the checklist had

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been added to the turnover sheets and noted that operators appeared to be adequately aware of the state of nuclear instrumentation during the remaining refueling activities in 1984. The inspector had no further question This item is close (0 pen) Unresolved Item (293/85-06-03). Core spray recirculation test valve MOV 1400-4A, o Following thislatestfinding(peratormountingcapscrewsfoundloos LERs83-010 and 83-035 report earlier problems)

tSe station again requested the Nuclear Engineering Department (NED),to evaluate the cause of the loose screws. The inspector discussed the re-commended corrective actions with the Station Chief Maintenance Enginee NED has recommended that a minor modification be made to the mounting screws (drill holes in the screw heads) and lockwire the four screws to prevent their coming loose. Previous efforts to use a thread compound and increased torque were not effective. The licensee is also continuing to evaluate system vibration. The licensee's planned actions include an inspection of the valve after each operation. The recommended modi-fication to lockwire the screws is pending availability to take the A-core spray out of service for environmental qualification upgrade re-placements. This item remains open pending further review of the lic-ensee's corrective action . Operational Safety Verification Scope and Acceptance Criteria The inspector observed control room o)erations, reviewed selected logs and records and held discussions wit 1 control room operators. The in-spectorrevIewedtheoperabilityofsafety-relatedandradiationmoni-toring systems. Tours of the reactor building, turbine building, intake structure, station yard, switchgear rooms, diesel generator rooms, bat-tery rooms, and control room were conducte Observations included a review of equipment condition, security, house-keeping, radiological controls, and equipment control (tagging).

These reviews were performed in order to verify conformance with the facility Technical Specifications and the licensee's procedure Findings (1) On April 5 1985 at 6:09 am the licensee brought the reactor criti-calfollowInganunplannedtripthedaybefore. The inspector ob-served the licensee s actions in the control room during a routine reactor vessel and reactor coolant system heatup to normal operating temperatures and pressure The inspector noted that the average rate of temperature change was within the T.S. limit of 100 F/hr and that the required temperatures were logged every 15 minutes as required by T.S. 4.6.A.1. However, the inspector questioned the Watch Engineer as to why the temperature logging was stopped at i

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4-J 10:45 am on April 5 1985 since the heatup of the reactor vessel metal was still in p,rogress and since T.S. 4.6.A.1 requires logging until the difference between two readings in a 45 minute period is less than 5 The licensee stated that the mode switch had been placed in the RUN mode, that the operators had'no control over metal temperature rise,

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and that pressure and temperature control were essentially in. auto-matic. The inspector concluded that'since the LCO only applied at vessel temperatures < 450 F and that the required logging was per-formed up to this point that the licensee's actions were acceptabl .At the exit interview, the licensee acknowledged the inspector's comments regarding confusion that some operators had because of the unclear ~ applicability of T.S. 4.6.A.1 logging requirement above 450 At the exit meeting, the licensee stated that a review of the T.S. wording would be performe o:

(2) On April 7, .1985 at 11:46 pm, the "F" channel of the average power

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crange monitors (APRM) tripped high, generating a half scram signa The licensee promptly cleared the half scram and determined that a momentary reactor' power spike of three to four percent had caused

'the "F" APRM to trip. The power spike was believed to be caused by the bistable vortex phenomena in the recirculation system de-scribed in NRC inspection report 50-293/85-03. The licensee stated that the. combination of a conservative trip setting and a low gain

' adjustment factor for the "F" APRM had caused it to trip before the

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other APRM channels. No further APRM trips of this type occurred during the inspection period. No unacceptable conditions were identifie (3) On April 12, 1985 at 11:55 am, the licensee determined that the re-fuel floor supply ventilation duct could not be isolated by second-ary containment dampers. Specifically, two dampers in series (AO--

N-82 and 83) would not properly close during a routine damper in-spection. The licensee initiated a reactor shutdown and notified the NRC via the ENS telephone line. The shutdown was secured at 1:46 pm when one of the two dampers was manually close The dampers were subsequently repaired and tested under maintenance requests (MR) No.85-282 and 85-283. The licensee replaced plastic drive gears on the dampers and repaired a broken damper blade. The

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dampers were returned to service on April 30,.198 Several secondary containment dampers, including A0-N-82 and 83, had some of their drive gears repaired during March 198 The licenseeindicatedthatthedamperproblemswererelatedtothe damper design. The secondary containment damper drive mechanisms are scheduled to be replaced during the next refueling outage with-

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dampers of heavier design.. The licensee indicated that the fre-

quency of damper visual inspections will be increased from quarterly to monthl e The inspector reviewed the secondary containment' leakage test re-sults for the current operating cycle and noted that the leakage

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has markedly increase In October,1984 the average differential

. pressure in the reactor building during a, test was -0.36 in. H,0 relative to outside atmos The next test, conducted on M8rch 18, 1985, demonstrated alower pher vacuum, -0.27 in. H 0. The licensee-could not explain the increased leakage during the 2second test and ,

' indicated that the next secondary containment leakage test would not be conducted until 1986,.during the next refueling outage. The inspector expressed concern over the length of the surveillance interval, in light of the change in secondary containment leakage and the recent damper problems. A subsequent licensee safety No. 1812, for a temporary modification of damper A0-N-evaluation,d 82 indicate that a secondary containment-leakage test may be ducted if future damper problems are foun The inspector discussed unclear aspects of the secondary containment-definition and.LC0 contained in_the plant's technical specifications with the licensee. The licensee indicated that the requirements in the standard BWR technical specifications were generally followed in utters relating to secondary containment. The inspector sug-gested that the plant's technical specifications be changed to bet-ter reflect the licensee's policy for secondary containment. The

, licensee acknowledged the commen The results of future damper inspections and secondary containment leakage tests will be reviewed during routine NRC inspections of the licensed program. The inspector had no further questions at this tim (4)

OnI&C an April techn17,ician accidentally shorted two test leads while perfo ing a calibration of local. range power monitors (LPRM). The test leads were shorted when'the technician inserted the leads into the current measuring test sockets rather than the voltage measuring test sockets on a digital multimeter during the tes The shorted leads damaged a recirculation flow converter unit for the flow biased circuits on the A, C and E average power range-monitors (APRM). The APRM's then tripped on a false high reactor power with no flow signa The technician indicated that he had been distracted by someone talking when he incorrectly connected the test leads to the multi-mete .

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6 The inspector reviewed the technician's work experience and con-cluded that the individual was qualified to perform the LPRM cali-bration. The incident appeared to be an isolated case of personnel

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error. The inspector had no further question (5) On April 24, 1985, the inspector noted that the "B" main stack radiation monitor was reading about 30% higher than normal. The licensee reviewed the stack strip chart recording and determined that the monitor output increased in a step change immediately fol-lowing a source chec<. The monitor subsequently returned to normal level The licensee checked both main stack detectors and concluded that the increase in the "B" monitor was due to an problem with the source check switch. The licensee stated that the problem was in-frequent and that the switch did not need to be repaired. The monitor functioned normally during the remainder of the inspection period. The inspector had no further question . Followup on Events and Nonroutine Reports Reports

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(1) system had been declared TheHighPressureCoolantInjection(HPCI)inspectionperiod(1:50 inoperable prior to the beginning of this am on March 31,1985) because of an overspeed condition caused by a broken electrical connector. Alternative system testing was in-itiated and continued until the HPCI system was returned to service.-

This connector was repaired and the HPCI system was tested on April 2,1985 but was not immediately returned to service because of in-dication of a ruptured exhaust line diaphrag Subsequently, a contractor working in the HPCI quadrant during the

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evening of April 4, 1985 noted a broken snubber shaft on the HPCI turbine exhaust line and immediately reported this to the control room. This snubber is in the overhead in the room and may have been damaged during previous operation and not notice Maintenance personnel as well as Quality Control personnel performed an inspection of the system. The station also called the Nuclear Engineering Department staff to the plant to evaluate adequacy of the piping support desig retested and reinstalle The inspector reviewed the licensee Thedamagedsnubberwasrebuilt,sevaluatIonofthi NED 85-338, dated April 6, 1985. This report included an evaluat on by the original designer, Teledyne Engineering Services, and recom-mendedrestorationofthepipesupport,toitsoriginaldesign(with different anchor bolts to ensure proper embedment because of being slightly pulled out of the ceiling). This report concluded that

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.a water hammer type event'had taken place and recommended short ters

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corrective actions to preclude further damage: 1) increase turbine -

exhaust line nitrogen purge time from 2 to 3 minutes after opera-o .

tion, 2) blowdown the line once per day, and 3) inspect the pipe clamp, snubbers, and baseplate after system operatio ,

The inspector noted that the::e actions were being implemented. The '

control room supervisor was keeping records of exhaust line blowdown

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done for 3 minutes each night and about midnight. The test proce-

.dureL(8.5.4.1) was revised to inspect the pipe clamp, snubbers and baseplate after system operation as well as perform a purge of the exhaust line for 3 minutes.

l The HPCI system was retested and declared operable at 1:20 an on l

' April 7, 1985. This event is described in .ER No. 85-08 dated April 26, 1985. This LER states that' additional root cause analysis is

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in progress and will be discussed in an update to this LER. The licenseeisalsore-evaluating 1)thetestmethodincludingadvan-ta j

s 2)ges the'and disadvantages capacity of vacuum of a quick relief on thestart" versus exhaust line.a slow start and The inspector had no further questions regarding this' event at this time. -A review of the licensee's root cause analysis and updated LER will be performed in a future inspection (85-08-01).

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At the exit meeting the inspector questioned the licensee's rationale for taking the reactor critical on April 5, 1985-(following an un-related main turbine tri being declared operabl onlicensee The April 4, 1985)

stated prior to the that the HPCI basis for system

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this included: 1) the demonstration that redundant safety systems l.'m wereoperable,iortoastartupfromaColdCondition,and3)the2)

be operable pr the req maintaining of the plant in a Hot Condition. The inspector verified L redundant system operability and maintenance of a Hot Condition and l had no further questions at this time, y

L '(2)~Anautomaticmainturbinetrip(andassociatedreactorscram)oc-L curred from 85 percent power at 11:04 am on April 4,1985 due to a high vibration signal from-a main generator bearing. Control room

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operators had earlier reduced power from 100 percent following re-

[ ceipt of a vibration alarm at 10:58 am.

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The inspector observed plant conditions shortly after the trip and b ,

noted that reactor safety systems responded normally and that no ECCS equipment was called upon to operate. Reactor coolant system pressure was 925 psig, all control rods were inserted, and operators were following the shutdown procedure (2.1.5).

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The licensee's investigation (with assistance of a G.E. turbine consultant) determined that the vibration sensing probe was not working properly (unevenly worn and insufficient oil flow). The vibration probe was replaced, a generator oil orifice was unplugged, and an unrelated broken selector switch for the "B" source range monitor was repaire The reactor was taken critical at 6:09 am on April 5, 1985 and the main turbine was placed back in service at 1:40 pm. The inspector had no further questions regarding this even Review of Licensee Event Reports (LER's)

Licensee Event Reports submitted to the NRC: Region I office were reviewed to verify that the details were clearly reported and that corrective ac-tions were adequat The inspector also determined whether generic im-plications were involved and if on site followup was warranted. The following reports were reviewed:

N Subject 85-04 Reactor Vessel Drain Line Leak 85-05 Missed Surveillance Test 85-06 Reactor Scram During Surveillance Test 85-07 Secondary Containment Dampers Inoperative 85-08 HPCI System Inoperable 85-09 Reactor Scram on Turbine High Vibration The events in LER's 85-04 and 85-05 were reviewed during Inspection N . The events in LER's 85-06 and 85-07 were reviewed during Inspec-tion 85-06. Additional failures of the secondary containment dampers described in LER 85-07, A0-N-82 and 83, are described in this repor The events in LER's 85-08 and 85-09 are described in this inspection re-port. No inadequacies were identifie . 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, Findings (1) TheinspectorwitnessedatestoftheHihPressureCoolantInjec-tion (HPCI) system at 1:00 pm on April 2 1985. This test was being performed to demonstrate system operabil ty following maintenance

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(replacement of an electrical connection on the turbine control equipment). The inspector noted that the repaired equipment oper-ated properly and that the test was conducted in accordance with the station test procedure (8.5.4.1). However, the inspector ob-served an annunciator that alarmed and was acknowledged by the operator during the test, "HPCI turbine exhaust diaphragm hi pres-sure". The control room operator and supervisor stated that this annunciator had been in alarm during a previous test, that two o)erators were on station in the torus room and the HPCI room to o) serve system piping and equipment, and that no reports of exhaust piping problems were observed, and there was no reason to imedi-ately secure the HPCI system (a system shutdown as discussed in the alarm response procedure, 2.3.2.2).

Following discussions with the inspector, the control room supervi-sor initiated a plant maintenance reguest and maintenance personnel inspected and replaced the exhaust piping rupture diaahragm. One diaphragm was deteriorated but investigation and chec<s indicated proper operation of the pressure switches for alarm and HPCI turbine trip. In kind replacement diaphragm rating was stamped as between 180.5 and 199.5 psig however, the HPCI turbine exhaust pressure tripwouldoccurearlier,atabout150psig. Exhaust pressure did not reach rupture pressure. LER No. 85-08 postulates that an ear-lier water hammer event may have caused the problem with the rupture disc. This LER is already being tracked (see IFI No. 85-08-01 in

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Section 5.a of this report).

The inspector had no further questions at this tim No additional problems were identifie (2) On April 17, 1985,ilation fueling floor vent radiation monitors.theThese inspector observed monitors gen- a cali erate a high level trip which isolates secondary containmen The licensee personnel performing the calibration followed the ap-propriate procedure, No. 6.5-170. However, the certification docu-mentation for the calibration radiation source they used was not complete. The licensee subsequently verified the calculated dose rates in the calibration source with an "R" chamber detector. The

"R" chamber had been calibrated by NBS. The licensee stated that additional documentation for the calibration source would be sought from the source vendor, General Electric. The inspector had no further question (3) On April 24, 1985, the inspector observed a portion of a functional test of the automatic initiation of the auto depressurization system (ADS)onhighdrywellpressure. The personnel performing the test

! followed the appropriate procedure, No. 8.M.2-2.1.6. No inade cies'were identified. The inspector had no further questions.qua-

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7. 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 implemente A list of the items reviewed is included at the end of this report in the attachmen Findings (1) On April 7, 1985, the licensee repaired the A0 220-44 reactor cool-ant sample line isolation valve. This valve had been repaired and successfully tested in March, 1985, but had subsequently failed closing time tests. The licensee indicated that the valve closing time was decreased on April 7, by modifying the nitrogen bleed off rate from the valve and by lowering the volume of gas in the valve-operato The nitrogen bleed off was increased by replacing an orifice on the solenoid valve that controls nitrogen flow from A0 220-44. The volume of nitrogen in the o)erator was decreased by lowering the nitrogen line pressure to tie valve from 50 to 25 psi The valve was subsequently tested and closed within the required time limit. The inspector had no further question '

(2) On April 22, 1985, the inspector reviewed the documentation for maintenance on the containment high radiation monitor. This monitor was declared inoperable on April 16, 1985 and was repaired and sub-sequently returned to service on April 19, 198 No inadequacies were identified concerning the maintenance wor However,indicatedthe 85-284, thatinspector noted had the equipment that been the maintenance request (MR) N returned to service prior to a supervisory review of completed post work testing. The post work testing was completed on a Friday, April 19, 1985, but the testing was not reviewed until the following week.

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The Chief Maintenance Engineer indicated that this incident was an isolated event, but that in the future, maintenance supervisors would be. called into the plant on the weekends to review lost work testing prior to closing out the maintenance requests. T1e inspec-tor had no further question , - - -- , -

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(3) On April 19, 1985, the inspector discussed. repairs to failed sec-ondary containment isolation dampers A0-N-82 and 8 The repairs to these dampers are described in Section 4 of this repor The inspector had no further questions at this tim . -Chemistry and Health Physics Activities On April 16, 1985, the inspector attended a meeting of the radiological oversight committee. A variety of topics were discussed at the meeting, including recent housekeeping problems in plant and the use of radiolo-

_gical occurrence reports. A number of recommendations from the indepen-dent radiological assessor were also discusse While the discussions were open and forthright, they were protracted and few decisions were mad The inspector subsequently expressed concern about the committee's ability to review the large number of Radiological Improvement few months. The Program licensee (RIP)indicated milestones-which willleted that the committee in the next be comp's efforts wou become more efficient as it gains experience in its oversight functio The inspector had no further questions at this tim 'The inspector reviewed the licensee's activities in preparation for a plannedfeedwaterHydrogenWaterChemistry(HWC)injectiontestatthe site in order to determine the need for any additional shielding of plant equipmen The inspector discussed system configuration and operation, and proposed radiological controls and surveys to be implemented during the tes The licensee has utilized the services of Advanced Process Technology Co., and the General Electric Co. to assess the impact of HWC at the station, to assist in test implementation, and take measurement The inspector determined that the licensee had included an assessment of both onsite and offsite radiological impact (including comparison with a regulatory dose rate restrictions). The licensee has also had special calibration studies done regarding survey instruments and monitoring de-vices response to the radionuclide involved. No problems were identified during this revie The following information is included in this report to assist NRC man *

agement in following radiation exposure at the station. The monthly 3 personnel radiation exposures for March and April, 1985 were 59.7 and'

78.1 person-rems, respectively. The total yearly radiation exposure through May 5, 1985 was 266.5 person-rems.

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The February monthly personnel radiation exposure was incorrectly re-ported in NRC Inspection Report 50-293/85-06. The reported exposure, 128.7 person-rems, was the yearly exposure through the end of Februar The correct monthly exposure for February,1985 was 52.2 person-rem No inadequacies were identifie . Followup on Licensee Review of the Hatch Unit One Vent Header Cracking Incident The licensee's initial response to the vent header cracking incident was re-viewed during NRC Inspection No. 50-293/84-01. During the current inspection period, the licensee response to a General Electric Service Information Letter (SIL) No. 402, "Wetwell/Drywell Inerting", issued February 14, 1985 was re-viewed. The Hatch event is attributed to the inadvertent injection of liquid nitrogen into a torus vent heade The response to this SIL is documented in a letter from Boston Edison Co. to NRC Region I, dated September 14, 198 The inspector verified the accuracy and completeness of the response by dis-cussing the SIL items with engineering and quality assurance department and by reviewing plant procedures and drawing No inadequacies in the licensee's actions were identifie Ont potential source of liquid nitrogen which could affect the primary con-tainment had not been fully evaluated at the time of the inspection. This source, the post-accident containment atmospheric dilution (CAD) system, is designed to deliver a relatively small volume (60 cfm) of gaseous nitrogen to the primary containment after an accident. The CAD system does not have low temperature cut off valves (present in other nitrogen lines) to prevent theaccidentalinjectionofliquidnitrogenintocontainmen The licensee stated that the nitrogen supplied to the CAD system would nor-mally be vaporized by truck-mounted vaporizers. The licensee plans to com-plete an evaluation of the effects of a failure of these vaporizers on the CAD system components and on the primary containment this summe The licensee evaluation of the CAD system will be reviewed during a future inspection (85-08-02).

10. 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 issuance. No written material was provided to the licensee during this inspectio _ _ _ _ - - _ _ . ___ ___- _ _ _ _ _ _ _ _ _ _ _ _ _ _-

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ATTACHMENT The following is a list of surveillance and maintenance items reviewed during this perio Portions of die following tests were reviewed:

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Closure test of reactor water sample line valves on March 28, 198 '

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Alternative system' testing between April 2 and 3, 1985 due to an inoperative HPCI syste Post maintenance full flow test on the HPCI system on April 2,1985(8.5.4.1).

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Routine calibration of the refueling floor ventilation system radiation moni-tors on April 17,1985(6.5-170).

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Routine functional test of the automatic initiation of the ADS on high drywell I pressure (8.M.2-2.1.6) on April 24, 198 Routine c'alibration of reactor high pressure sensor PS-261-23A on April 22,

-198 Also reviewed calibration history from 1981 to April 198 Portions of the following maintenance items and temporary modifications were reviewed:

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M.R. 84-1-55 Main steam line high flow switch replacement, environment quali-fication upgrad ,

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M.R.85-284 Repair containment high radiation monito M.R.85-282 Repair damper A0-N-8 M.R. 85-28 Repair damper A0-N-8 M.R.85-287 Repair flow comparite M.R.85-267 Replace SV-220-4 l

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M.R. 85-45-107 APRM meter 3% lo M.R.85-289 A0-220-45 limit switch not workin Temporary modification 85-22, remove demisters from the standby gas treatment syste '

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Temporary modification 85-25, A0-N-82 drive louver repai _ - _ _ _ _ _ _ _