ML20128E055
| ML20128E055 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/16/1974 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20128E027 | List: |
| References | |
| NUDOCS 9212070460 | |
| Download: ML20128E055 (37) | |
Text
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f SUPPLEMENTAL REPORT ON Tile OPERATION OP THE NORTilERN STATES PO'..'E!! COMPANY MONTICELLO REACTOR FACILITY Prepared By The DIRECTORATE OF REGULATORY OPERATIONS NORT18FRN STATM POUEP. CO.9.".!P!
(MONTICELLO - DOCKET NO. 50-263)
IN CONNECTION WITil FULL TERM OPERATING LICENSE APPLICATION October 16, 1974 9
9212070460 741106 PDR ADOCK 05000263 P
t' TAlt!I Ol' CONTENTS INTRODUCTION 1
2
SUMMARY
3 CONCLUSION 4
I DISCUSSION 1.
Ol'ERATING ll1 STORY 5
11.
UNUSUAL OCCURRENCES 5
A.
Reactor and Auxiliary Systems 5
1.
Reactor Scram and Subsequent Safety 5
and Relief Valve Maloperation (7/72) 2.
Reactor Scram and Subsequent Relief 6
Valve Maloperation (7/72) 3.
Main Steam Isolation Valve Leakage 6
(Spring 1973 and 1974) 4.
Failure of Two Main Steam Isolation 6
Valves to Close (2/74) 1 B.
Emergency Core Cooling Systems 7
1.
liigh Pressure Coolant Injection 7
System Failures (7/72, 5/73 and 5/74) 2.
Residual llent Removal Pump Motor 8
Failure (12/72)
C.
Containment 8
1.
Torus-Drywell Vacuum Breaker 8
Maloperation (2/72) 2.
Primary Containment Leakage (5/74) 8 D.
Effluent Systems 9
liydrogen Detonations in off-gas System 9
(5-7/74)
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a 111. RADIDACTIVE WASTE DISPOSAL 9
A.
Caseous Effluents 9
B.
Liquid Effluents 10 C.
Independent :teasurementr. of Plant 10 Releases by Regulatory Operations IV.
ENGINEERED SAPEGUARDS 10 A.
Containment 10 i
B.
Emergency Elc-trical Power System 11 C.
Emergency Core Cooling Systems 11 1.
liigh Pressure Coolant Injection 12 System 2.
Low Pressure Coolant injection and 12 Core Spray System V.
SAFETY SYSTD1 PERTOWANCE 12 A.
Reactor Saf ety Systera 12 B.
Reactivity Control 13 C.
Reactor Pressure Relief System 13 e
VI.
PRIMARY SYSTEM INTEGRITY 14 J
VII. RADIATION PROTECTION 14 A.
Staffing and Training 14 B.
Procedures 15 C.
Cencral Plant Cleanliness and Control 15 D.
Personal Radiation Exposure Control 15 E.
Monitoring and Counting Systems 15
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V111. ENVIRON.'! ENTAL f t0N170RI:;G/121ERCC:;CY PLANN1!40 15 IX.
VIOLATIONS 36 l
X.
OPERATING ORGANIZATIO::
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j Attachmentst Table 1 - Summary of Regulatory Inspections Table II - Chronology of Operation Table III - Reactor Scrams
-Table IV - Violations T
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. t SUPPLD;E;TAL REPORT LY Tilt DIRECTORATE OP RECUL\\ TORY OPERAT10!;S OPEPATION OF Tile !!O!;TICELLO Nt' CLEAR CE?;EPATit:G P wit U;; DER PROVISIONAL OPER.\\ Tit ; 1,1CP.;:SE DPR-22 INTRODUCTION The Directorate of Licensing issued a safety evaluation for full-tern 11 cense review of the Monticello liuclear Plant on February 5, 1973.
Included as Appendi>; A to this tafety evaluation was a special Summary report issued by the Directorate of Regulatory Operations on August 4 1972, which described the cetivitico of the Monticello !!uclear Plant through May 1972, as observed by the tegulatory Operations inspection program and discussed in various licensee reports. This supplen. ental report provides an updated sunnary of subsequent safety-related plant activities through June 1974.
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SU:DtARY The Monticello Nucicar Generating plant has completed an additional 25 raonths of commercial power operation sinco our initial sut. mary report van issued.
During this period the plant has generated an additional 7,022,020 Mw-bra of electrical energy, raloing the total energy output of the reactor since initial criticality to the equivalent of 36,771 hours0.00892 days <br />0.214 hours <br />0.00127 weeks <br />2.933655e-4 months <br /> of full power operation The most significant problems experienced since May 1972 have been associated with relief valve and nain steam isolation valve operation, primary containment vacuum breaher and isolat.f on valve Icakage, high pressure coolant injection system operability, and several individual component failures as discunced later in this report.
The overall performance of installed control and safety systems has been essentially as designed. The licensee's surveillance tenting program has proven to be effective _in that most operating problems experienced have been discovered by scheduled tests or innpections.
The management, operating, and engineering support organizations have remained stable since our initial report was issued, c:: cept for transfer from the plant staff of two supervisory engineers who were replaced by experienced engineers from the staff.
A special inspection of management controls in May and June, 1972, identified several matters which prompted improvements in the licensee's management programs.
He consider these improvements to have shown effective results.
Review, investigatiou, and resolution of operating probicms have shown improvement and are generally thorough as shown by the small number of recurrent failures.
We judge the operation of the Monticello plant and the competence of the plant staff to be adequate to assure continued safe operation.
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CONCLUSION The findings of our inspection program since our initial special summary report was issued show that the 11ontiecllo Mucicar Generatint, Plant has been operated safely since initial startup in 1970.
We find that the reactor and its control and safety systems have continued to operate as designed except for main steam isointion valve leakage (which is currently showing an inproving trend) and ludividual component failures which have 1
been corrected.
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DISCUSS 10:1 i
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Review of operation of the Monticello liucicar Generating Plant by the Directorate of Regulatory Operations has continued on a regular bant.1 since the initial r.ummary report was issued. This review was accottplished through a total of 22 inspections in addition to those listed in the initial report, representing an expenditure of approxinately 97 tnan-days at the plant site or corporate officet,.
A listing of the additional Inspection dates and general arcan inspected are given in Tabic 1.
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in addition to these inspections, frequent informal contact with the licensee's organization has continued along with regular review of
,i operating, abnortaal occurrence, and other reports submitted to the AEC i
l by the licensee.
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t The significant findings of our inspection program are discunsed and evaluated t follows:
1.
Operatine. History Operation of the Monticello plant has continued with only brief interruptions since May 1972 except for scheduled refueling outages in the spring of 1973 and 1974.
A chronology of significant events related to plant operation is given in Table 11.
The number of scratas which have occurred since the start of comnercial operation (July 4, 1971) and other pertinent operating statistics are sunmarized below. The dates and associated causes of reactor scrans since May 1972 are listed in Table III.
No. of Times No. of Equivalent Gross Brought Reactor Tull power E1cetrical Year Critical Scrann Hourn Mv-hours.
1971 (beginning 7/4) 11 7
2131 1,241,0S0
-,717,750 3
1972 20 8
6537 1973 8
3 5999 3,412,160 1974 (thru June)
, 5, 2
2104 1,196,780 TOTAL 44 20 16,771 9,567,770 11.
Unusual Occurrences This section suntarizca the more significant occurrences associated with the operation of the reactor subsequent to May 1972.
A.
Reactor and Auxiliary Systems 1.
On July 10, 1972, with the plant operating at full power, a loss of generator excitation resulted in a reactor scram and isolation.- The D relief valve did not lift as required, and the A relief valve lifted but did not resent until reactor pressure had decreased to 600 psig.
The operabic relief valves maintained reactor pressure within prescribed limits.
The A safety valve also opened briefly, and raised dryvell pressure to 2 psig. Althour' subsequent investigation showed that two of the four l
instrument taps which sense drywell pressure were covere-with tape, the two redundant instruments initiated e=argen:y core cooling systems, which functioned as required.- Vollow-ing the event the _ plant was placed in a cold shutdown condition for investigation.-Inspection by the licensee showed that 1
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rust particles had prevented prompt rescating of the A relief valve. No reanon for failure of the D relief valve to open could be determined.
Review of the history of the A safety valve gave no reason to suspect its having operated inproperly. After cleaning of A relief valve components, inspection of the B and C relief valves, and satisfactory te. sting of all relief valves the plant was returned to power. A report of the occurrence was submitted to the ALC by SSP on July 20, 1972.
2.
On July 21, 1972, spurious initiation of the main transforner fire protection system resulted in a scram from full power, and the D relief valve again did not operate.
All other sy=tems functioned as required. The plant was placed in a cold shutdown condition for investigation.
Further inspection of the D relief valve revealed a small leak in the bellows and a design deficiency which prevented the leak detection system from detecting a small Icak.
The Icaking bellows was replaced, other relief valve bellows were inspected, and the Icak detection system was, modified to facilitato detection of small bellows leaks in the future.
The occurrence was reported to the AEC by NSp in a letter dated July 28, 1972.
3.
Scheduled lechcge tecta conducted during the 1973 refueling outage showed four of the eight main steam isolation valves (MSIV's) to be leaking in excess of Technical Specifications limits. A truing cut was made on the main discs, additional stellite material was added to one seat, and the seats on the four valves were icpped.
Subsequent tests showed satisf actory leak test results.
A report of-the MSIV leakage was submitted to the AEC by NSp on June-28, 1973.
Subsequently, during the 1974 refueling outage, one of the eight MSIV's was observed to leak in excess of the amount allowed by Technical Speci-fications. A truing cut was made on the main and pilot valve plugs and both seats were lapped. -The leak rate during a subsequent test was satisfactory.
This experience was reported to the AEC by the licensee in a letter dated May 20, 1974.
4.
During a routine surveillance test conducted on February 16, 1974; two of'the eight main steam isolation valves did.
not close as required. -Acredut. dant MS1V in series with each of these two functioned' properly.
Investigation showed that the rubber-seats in the solenoid valves associated with new MSIV operators installed during the-1973 refueling outage -(see section IV. A)- had deformed in-
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such a manner that normal operation of the solenoid l
valves did not occur.
The seat s were replaced with spring-loaded ceats of an improved design to prevent recurrence.
The occurrence was report ed to the AEC by the licensco on February 25, 1974.
In two other instances an MSIV required in excess of the allowed 5 seconds to close, due to miralignment of supporting rollers which ride against the yoke rods of the valve.
The roller arrangement was modificJ during the 1974 refueling outage to prevent recurrence.
These events were reported by the licensee in letters to the AEC dated August 10, 1973, and March 15, 1974.
B.
Emercency Core Cooling System 1.
Four failures of the high precoure coolant injection (HpCI) system have been experienced,.f rem unrelated causes, since May 1972.
During a test on July 17, 1972, the HpC1 turbine tripped because of high steam exhaust pressure.
Investigation showed that the disc in an exhaust line check valve had detached and was partially blocking the exhaust line to the torus.
The disc was reinstalled using a sturdier dice pin.
On July 31, 1972, the UpCI turbine control valve did not respond properly during a test.
Investigation showed pieces of a plastic pipe cap in the governor oil system.
The plastic f rant.ents were removed, and portions of the governor system were disassembled and inspected. The licensee concluded that the pipe cap had been introduced into the system during plant construction.
During a " quick start" surveillance test on May 18, 1973, the HpCI system isolated because of apparent high steam flow.
Investigation by the licensee disclosed an intermittent electrical circuit in the control system, worn drive gears and a loose coupling associated with the electro-hydraulic acturator, and a loose gear associated with the turbine speed signal to-the control system.
The loose parts were secured with scif-locking set screws, the worn gears vere replaced, and a-related oil passage was modified to provide more effective lubrication of the gears.
During a post-maintenance test on May 21, 1974, the HPCI auxiliary oil
_ pump did not start because of a misaligned contact assembly associated with the oil pump motor acceleration relay.
A post-maintenance inspection of the relay was incorporated into the maintenance procedure to prevent _ recurrence.
In each of the four cases a satisfactory postrepair test of the-system was conducted and recurrence has not been I
experienced. The failures were reported to the AEC by the licensee in letters dated July 27 Jand August 3, 1972, May 24, 1973 and May 30,1974,.respectively.
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2.
A residual heat renoval (P,11R) pump motor developed a ground fault during operation on December 16, 1972.
Investigation showed the failure to have resulted from insulation damage caused by cracking and movement of the lower air deficctor.
Ncw air deficctors were installed in all f our pJm pump rantors using an improved method of attachment to prevent recurrence of the cracking.
The occurr(nce va reported to the AEC by NSP in a letter dated December 22, 1972.
C.
Centainment 1.
During a scheduled inspection of the torus-to-drywell.
vacuum breakers in December 1972, one of the ten vacuum breakers was found to be approximately 1 1/4 inch open, although it was indicated to be in the closed position.
An operational test of the. vacuum breakers was then performed,-and ivur did not fully close.
The shnft seal was modified on all ten vacuum breakers to allow free operation. During the 1973 refueling outage.the ascociated position indicators were changed to be sensitive to movement of approxicately 1/16" from the full closed position, and, the manual actuating arms were-relocated to provide a higher closing torque near the fully closed ponition.
An ennuncintor syatee uns 9ubsequently installed to warn the operator when a vacuum breaker is not fully closed. The occurrence and-followup actions were reported to the AEC by NSP in letters dated December 22, 1972 -and March 12, 1973.
2.
While performing' an integrated primary containment leak rate test in ?!ay 1974, the licensee measured a leakage rate of approximately 3 percent per day.
The operational limit is 1.2 percent _per day. 7 Investigation by_the licensee-showed that an air-line to a torus-to-dryvell vacuum breaker test operator (located inside the torus) had been left open when the vacuum breaker ~and its operator were _ removed in September 1973 for dynamic testing.
This created a Icakage path aince the three-way solenoid valve installed in the line outside the primary containment is-normally positioned to vent the air supply line to the reactor building. The air _line was capped and to provide a positive isolation at the; containment-boundary, a i
manual block valve-was installed in each air supply _line.
The occurrence was reported to the AEC by the-licensee on May-24,-1974.
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D.
Effluent Systemn i
i During startup testing of the newly installed off-gas holdup i
system recombiners in thy 197a, a hydrogen detonat ion occurrt d
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in the recombiner inlet piping.
A flow control valve in the ree,nbiner was believed to have nonerated a spath whf eh initinted j
the detonatien, and the internals of four control valves were replaced with unn-cparking materials.
During subsequent 1
startup testing in June 1974, a accond detonation was experienced.
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The licensee made further modifications to the system to j
minimize the possibility of detonation and installed special 1
instruments on the system piping to provide data which could 2
assist in the eva3untion of any subsequent detonation. Gaseous activity was released from the reactor building vent exhaust l
af ter each detonation, although the amounts were well within license limits, t.dditional details concerning the first two 1
detonations were included in licensee reports to the Directorate 4
of Licensing dated "ay 29 and June 20, 1974.
Following restart i
of the recombiner system in July, 1974, a third detonation occurred.- No release occurred, since the air ejector rupture j
discs had been removed following the-second detonation. The j
system is designed to withstand the pressures generated by a hydrogen detonation, and inspection by the licensee af ter each j
of the three detonations showed=no damace to-have resulted.
l Data recorded by the rpecial instrumentation enabled the licensee to determine the location of the detonation and led him to the conclusion that the detonations had occurred as a 4
1 result of catalytic recombination of hydrogen and oxygen in the recombiner inlet piping.' Activation analyses showed traces of recombiner catalyst to have been introduced into the inlet piping by a system flush performed during'the spring of 1974. The licensee is evaluating methods of removing or poisoning the misplaced catalyst material.
III. Radioactive Waste Disposal A.
Gaseous Effluents The licensee has installed an off-gas holdup system which is designed to reduce gaseous effluents from the plant by a factor of 100.. This system'is being tested and is expected to be placed in service during the f all of 1974.
Gaseous effluents released to the environment during installation of the modified system have remained below the applicab)e AEC license' limits.
as shown by the following data e
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i 1972 1973 1974*
Type Discharge Curies
% of Limit Curies
% of Limit Curies
% of Li-it i
Noble Cases 751,000 8.8 869,000 10.2 747,000 17.6 Halogens 0.576 )
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5.3 Particulates 0.0125 )
0.009 0.029
- January through June j
B.
Liquid Effluents _
i The licensee has continued to_ place emphasis on liquid waste j
recycling to the extent that no radioactive liquids have been j
discharged from the Monticello plant since January 4, 1972.
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Independent Measurement of Radioactive Effluents By
. Regulatory Operations i
Our independent measurements 'of radioactive effluents continue to show results consistent with.noasurements reported by the i
licensee.
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Engineered hafecuards A.
Containment a
l Containment integrity was successfully demonstrated during.
~ integrated primary containment leak rate' tests conducted at the end of the spring 1973 and spring 1974 refueling outages.
Individual penetration Icak rate tests conducted at the beginning of these outages identified several containment
-isolation valves to-be-leaking in excess of allowable amounts.-
-These were all satisfactorily retested following maintenance.
The number of valves found to be leaking decreased between 1973 and 1974, indicating an improving trend as a result of the maintenance perforced. The leaking valves included four main steam isolation valves (MSIV's) in 1973 and one in 1974, as further discussed in Section 11.A.3.
Modifications were made to the MSIV's and their air operatoro during the 1973 refueling outage to correct operational problems experienced during previous months, as discussef in a letter f rom the.
licensee to the Directorate of Licensing dated January 22, 1973. This letter noted that execpt_for one slow closure-(_ discussed in'Section II.A.6 of our initial Report.of: Plant
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OperationF, the MSIV operator problems had not affected the
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ability of the valves to close when required. The modifications performed during the 1973 outage resulted in improved MSIV operation, although two MSIV's subsequently failed to close during a surveillance test in February,1974, as further discussed in Section II.A.4.
Several licensee reports in late 1972 and early 1973 discussed Icakage problems associated with-primary containment. atmospheric control valves which use a pressurized rubber seat.
Design changes were made to these valves during the 1973 refueling outage and a surveillance program was established to monitor their performance.
One of them leaked slightly in excess of the allowabic amount during a 1974 local leak rate test because of a scale deposit on, the seating surface, and a similar valve was found to have a linkage misalignment in November, 1973, but the rubber-seated valves have otherwise operated without reported failures since July, 1973. A prob 1cm with torus-to-drywell vacuum breaker operation and the discovery of a primary containment leak are discussed in Section.11.C of this report.
The pressure switch which operates'one of two reactor building-to-torus vacuum breakers failed in November, 1972. A~new bellows assembly was installed. We note that the containment-related problems discuased in this paragraph were all discovered as a result of inspection or sutveillance performed by the licensee, and were reported to the Commission as required by Technical Specifications.
In each case, prompt corrective action was initiated by the Itcensee to prevent reevrrence.
B.
Emergency ~ Electrical Power System Our inspection findings show the diesel-generators to have been functionally tested as required since our initial report was issued except for one inadvertent omission which was detected and reported by the licensee.
The licensee also reported three separate failures of an individual diesel-generator starting system, but because of thel installed redundancy none of t1.ae events rendered a diesel generator inoperable.
Except far planned maintenance:or surveillance as authorized' by the Technical' Specifications, both diesel generators have been available for use at all times.
C.
Emergency Core' Cooling Systems (ECCS)
L The inspection findings of our. original report remain unchanged;-
that is, that the installed redundancy of ECCS subsystems has proven to be adequate in that ECCS operability 1ms continued to meet,or_ exceed Technical Specifications requirements.-
Principal problems experienced since our initial report was issued are. discussed below:
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Operability tests of the liigh Pressure Coolant Injection (llPCI) subsystem have continued at monthly intervals.
Inoperability of the llPCI' subsystem was reported on four occasions, as discussed in Section II.B.1 of this report.
One other IJcensee report discussed an improper pressure switch setpoint which would have prevented operation of the llPCI subsystem as required between 150 and 167 psig I
reactor presaure.
The redundant ECCS components were operabic during each of these conditions, and prompt corrective actions were taken by the licensee in each case.
2.
The licensee has continued to perform monthly operability tests of the low pressure coolant injection.(LPCI) and core spray subsystems.
During one of these monthly tests in April 1973, one LPCI injection valve could not be opened using the control room hand switch because of
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dirty auxiliary contacts.
The valve operated properly after c1 caning of the switch and no recurrence has been j
reported. Broken flow switch paddles were found on residual heat removal (R11R) pumps No.11 and 13 in August, 1972 (the Rim pumps serve as LPCI pumps when LFCI subsystem operation is required)..The-paddles were modified and a safety analysis showed that the missing parts could l
present no ha?nrd to the reaeter. A ground f ult en ene RilR pump motor occurred during pump operation in December, 1972, as discussed in Section II.B.2.'The licensee also reported in March,'1974, that one ECCS-initiating reactor vessel level switch failed to trip during a routine surveillance test, although three redundant switches would.have initiated the ECCS if required. A new switch was installed and satisf actorily tested.
In each of the l
circumstances described, redundant.ECCS equipment would i
have provided proper core cooling if required. No failures related to core spray subsystem operation were reported or detected during the conduct of our inspection program, i
V.
Safety System Performance A.
Reactor Safety System Our inspection findings show that the reactor safety system has continued to operate as designed.
The licensee reported that (1) a small-Icak was found in one of four condenser j
- ' ~J vacu i scram switches (which introduced a conservative error) and (2) one of sixteen -steam tunnel high temperature switches i
vas:found to have a setpoint slightly (4*F) above the Technical-
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Specifications limit. No other safety system problems were reported and-neither of'these would have prevented a reactor scram if required.
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Reactivity Control The control rod system centinues to perform satisfactorily in that a control rod has never failed to insert when required. The licensee reported in June, 1972, that insertion of one contrv'. rod stopped at the "02" position (six inches f rom fully inserted). This has since been observed en other occasions, although'the licensee-demonstrated in each cane that_ scram insertion times were within J
Technical Specifications requirecents. The licensee and reactor j
supplier attribute this behavior to normal wear of piston seals, and note that the reactivity effect of the last six luches of rod insertion is uinimal.
Special tests performed in 1973 on five control rods observed to have stopped at the "02" position verified the observed beaavior not to be due to mechanical
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interaction with other core components. Our findings are that this behavior has had no effect on the ability of the control
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i rods to saf ely shut down the reactor.
During the initial startup following the 1973 and 1974 refueling outages, the licensee noted differences between predicted. and actual-critical rod pattern.
In 1973 the difference was due to prediction inaccuracies. The difference in 1974 was due to underesticated reactivity worth of the gadolinium in the' newly inserted fuel bundles.
In each case the licensee followed= appropriate precautions while determining-the reasons for the difference and incorporated improvements into his prediction techniques.
Testing during the 1974 refuelius outage resulted in replacement of si:, control rods due to inccrted neutron absorber tubes. during manufacture, although shutdown targin tests concacced in August, 1973, had shown the contrci l
rods to provide a conservatively adequate shutdown capability, i
i C.
Reactor Pressure Relief Sys' tem The pressure relief system has continued to demonstrate its ability to protect the reactor from overpressure, although probicos with safety and relief valve operation were observed on 4
two occasions in July,1972 (see Section II. A.1 and II. A.2). One j
other case of relief valve sticking occurred during the= initial test (at reduced reactor pressure) following the 1973 refueling outage. - The licensee deteruined this problem to have resulted from a bent' air operator stem. The: condition was corrected and' no subsequent relief valve problems have been reported.. The.four original safety valves (designed to relieve steam directly to the-drywell) were removed during the 1974 refueling outage.and four additional. relief valves were installed Oith discharges piped to the bottom-of the suppression pool.
This provided additional margin for proper pressure-relief capacity, since six of the eight installed relief-valves will provide adequate relief capability.
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primary System Integrity Our inspection findings are unchanged.from our original report; that is, that Icakage from the primary system has been limited to
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leakage through valves, valve packings, and pump seals. Eincteen of twenty studs which secure the stuffing box holddown cover en the two recirculation pumps were discovered during the 1974 outage to have cracked, with one stud having sheared completely.
- o loss of primary systen integrity resulted.
Investigation by the licensee showed the cracking to be due to improper heat treatment. To prevent recurrence, the studs vere modified'to eliminate an internal cavity (which caused accelerated corrosion), monitoring of material properties was increased, and replacement studs were provided.
Another complete set of steds with more rigid material specifications was also ordered.for installation during the-1975 refueling outage.
4 An inservice inspection program for the primary system boundary, as required by Technical Specifications, was initiated during the 1973 refueling outage and' continued during the 1974 outage. So abnormal j
conditions have been reported by the licensee as a result of these inspection activities.
The licensee has completed, at our request, a program to determine that valves important to nuclear safety have the required wall thicknesses. All valves examined during the conduct of this program were found to be acceptabic. The licensee has also installed additional instruments to monitor and record primary syctem 1 :h:g:. L'cing inputs to the prece: 00mputer, th:
new system is designed to give an indic: tion of instantaneous lash rate when desired and provides an alaru if a Technical Specification i
Icak rate limit is reached.
VII. Radiation Protection i
A.
Staf fing and Training The radiation protection staff consists of trained and experienced
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supervisory personnel and health physics technicians.
Staff turnover has been minimal and has not involved-_ supervisory personnel. During outages,-radiation protection staffing assistance has been obtained:as necessary from another nt: lear facility within the company and by contract from.another organization.
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All persons-coming to the facility for work assignment are provided continuous escort in restricted. areas of the facility until the individual has successfully completed the plant's indoctrination.and training program..Out-inspection findings.
have shown these training measures to be effective.
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Procedures r
The licensee has a written " Radiation Safety Manual", which i
defines responsibilities, training, radiation protection policies, and various procedures and practices.
This manual is reviewed and updated on a continuing basis.
L C.
General Plant C1canliness and Control-Our inspection findings show the licensee to be effectively maintaining plant cleanliness to facilitate control.of radiation exposures.
In addition to regular decontamination practices, i
the licensee uses plastic shecting, access control, and other measures to prevent the spread of contamination.
Individual whole body counts performed at the end of each refueling i
outage have not shown a significant intake of radioactive material and have detected littic radioactivity of plant 4
origin.
D.
Personal Radiation Exposure Control d
Personal exposure to external radiation is monitored by use of thermoluminescent dosimeters.
In addition, when entering areas where significant radiation levels exist, personal monitoring is performea on a daily basis ey issuance and daily reading of pocket dosimeters.
No personal overexposures-have occurred. Total whole-body exposure at.Xonticello in 1973 was 154 man-rems. The highest individual exposure f or 1973 was 4.3 rems.
1 E.
Monitoring and Counting Systems Our inspection findings show that installed monitoring and j
counting systems have been effectively used and maintained by the licensee.
VIII. Environmental Monitoring / Emergency Planning-During an August, 1973, inspection it was noted that environmental i
sampling onmissions had occurred. The licensee responded to-inspection-findings by initiating program improvements and corrective actions
+
to prevent further ommissions. _The Monticello radiological environmental l
sonitoring program complies with currect AEC regulatory requirements.
The licensee's driginal Emergency Plan, issued in 1970, is soon to be revised to conform to the requirements of the subsequently g
t-issued Appendix E to 10 CFR 50.
A draft revision was issued in 4
1973 for review and comment by State acencies and other offsite participating groups.
The -licensee plans to issue the revised Emergency Plan in final form in the near future, af ter comments i
"from participating groups are resolved.
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1X.
Violations I
Several violations of license or regulatory requirements (previously termed " noncompliance itens") have occurred during the period covered by this supplemental report.
Fifteen viciations were identified during a special inspection of management systems conducted by a team of Regulatory Operationn inspectors in : lay and June, 1972.
Our regular inspection progran has recorded twenty other-violations (through June 1974), nine of which were discovered by the licensee and reporced to the AEC.
The twenty violations are 4
categorized as follown:
A Limiting Condition for Operation 6
Surveillance Test Requirements 8
Quality Assurance Program 3
other 3
A listing of all violations identified during the period of this report and the related corrective actions is attached as Table IV.
t X.
_ Operating Organization i
The basic organization of the plant staff remains as described in our initial report.
Encept for promotion of the Plant Engineer Technical arid Plaat Ein;ineer Operations cv viisite jobs, no manage-ment or supet:isory personnel changes have occurred. These two vacancies were filled by experienced engineers previc.usly assigned
{
to the plant staff. The Monticello plant staff has demonstrated itself to be competent to assure safe plant operation. Organi-i zational changes made in early 1972, improvec management programs, and additional experienced gained in plant operation have further strengthened the licensee's technical support capabilities.
Problem j
review and investigation have generally been observed to be thorough and effective. A special inspection of management systems conducted in May.and June of 1972 identified several-areas-in which improvements in management control could be made (see'Section IX)..Some of the identified areas were noted to have been corrected by the early 1972 reorganization of-the plant staff. Additional improvements in management systems prompted by the inspection findings are considered to have been effectivo in providing the required review and control of-plant operation.
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Table 1 - Summary of Reculatory Innpections Dates of_1nnfection_
Princ@ ni Activities Reviewed 6/7-8/72 Corporate !!anagement Systens.
7/12/72 Renetor Scran, railure of D relief valvc to operate, and related events.
10/3-5/72 Plant operation and maintenance.
11/28-30/72 Plant operation and naintenance.
1/30/73 -
Plant operation, corrective actions related 2/1/73 to May 7? nanagement inspection.
2/20-22/73 Corrective actions from nanagement inspection, outage plans, radiation protection.
2/27/73 -
Emergency planning.
3/1/73 3/27-29/73 and Refueling and maintenance activities, in-service 4/4-5/73 inspection, valve wall thickness verification proccam.
5/23-25/73 Refueling outage activities, plant operation and maintenance.
5/29-31/73 P.adwaste systems.
6/27/73 Physical Security.
7/1" 73 Nuclear !!aterial Safeguards.
'7/i.7-19/73 Plant operation and maintenance, corrective
'h actions from management inspection, off-gas g3.w system testing.
8/30-31/73 Environmental monitoring program.
10/24-26/73 Plant operation and maintenance, corrective actions from management inspection.
12/18-20/73 Plant operation and maintenance.
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Dates of increction Principal Activities Reviewed 3
l 12/26-28/73 Emergency planning, environmental nonitoring.
l 2/27/74 -
Radwaste Systems i
3/1/74 1
l 3/5-8/74 Plant operation and maintenance, refueling l
preparations.
I 3/26-29/74 Quality assurance manual, plant maintenance and refue)ing activities, radiation protection, in-service inspection, relief valve installation.
5/9/74 Radwaste systems.
6/18/74 Relief valve installation.
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j Table II - Chronology of Operation i
2 6/1/72 Operating at 100% power.
4 6/2/72 Scheduled outage to replace "A" safety valve and perform plant maintenance.
Returned to operation 6/4/72.
6/5/72 operated at 100% power except for brief reductions for to valve exercising and brief turbine-generator outage on 7/9/72 6/1/72.
During this period No. 11 and No. 13 RHR service water pumps failed to meet Technical Specifications head-flow requirements.
Corrections were made to indications, and a Technical Specifications change was subseouently issued.
Inspection of torus ring header during this period revealed damaged bolts and improperly j
cut holes.
7/10/72 Reactor scram caused by faulty connection in generator i
amplidyne control circut.
"A" safety valve opened v momentarily below its setpoint.
"D" relief valve'did not open and "A" relief valve.'did not rescat properly.
Two drywell pressure sensing taps were found to be covered with tape, apparently from plant construction.
Reactor remained in cold shutdown'during repait and investigation.
"A" and "D" relief valves were inspected 4
and tested. All-containment pressure sensing taps-were verified to be clear.
1lork was done-to eliminate improperly cut holes on torus ring header. Returned to operation-at low power _on 7/16/72..
i 7/17/72 Increased to and operated at 100% power. On 7/17 the to HPCI turbine exhaust check valve disc pin f ailed; the disc 7/20/72 blocked the exhaust line, causing the rupture discs
-to burst.
7/21/72 Reactor scram caused by spurious initiation-of main-transformer deluge system.
"D" relief valve again did not open.
Investigation showed small leak in pilot bellows. Bellows leakge monitoring system was modified to provide sensitivity to small leaks. -Main transformer bushing was replaced.
7/31/72 During reactor heat-up, HPCI calfunctioned due.-to
-foreign material in-hydraulic system. Foreign material.
l-was removed and HPCI was tested.
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8/2/72 Returned to power operation, operated at 100% except to for brief reductions for valve exercising.
Broken 9/22/72 flow switch paddles were found on No. - 11 and 13 RHRl pumps on 8/31.
Paddles were modified to prevent:
further loss of material, on 9/15, one starting system I
'on No. 12 diesel generator operated slowly due to a plugged' air relay orifice. Air relay orifices for all o
starting systems were cleaned.
9/23/72 Scheduled shutdown to change control rod sequence, Satisfactorily test operated all relief valves while i
shut down.
9/24/72 Resumed power operation.
Increased to.and operated at to 100% power except for brief: reductions for valve 11/22/72 exercising.
On 10/11, No. 12 standby liquid control pump operated improperly due'to air in suction piping.
Procedure was modified to prevent recurrence.
11/23/72 Reactor power reduced to 85% due to ice blockage of circulating water intake. Normal deicing line not usable due to silting and inoperable discharge gate operators.
Placed cooling towers into operation.
Resumed operation at 100% power, i
11/27/72 Reduced power to dredge river and repair. discharge gate j
operators.
Removed cooling towers;from service.
f 11/28/72 Resumed operation at 100% power.
I 12/15/72 Scheduled outage to clean condenser and perform I
to maintenance and inspections. Testing of torus-12/20/72 drywell vacuun breakers disclosed sticky operation.
i Stem seal was modified to-provide proper operation.
Ground fault occurred on No. 11 RER pump motor.due to j
failure of. air-deflector.
Cracked air deflectors were also found in No. 12 and 13 RHR pump motors.
Outage was extended to-repair and modify all RHR pump motors.
Several MS1V spool valve malfunctions occurred, necessitating cleaning or replacement of spool valve assemblies.
I 12/21/72 Continued' operation at 100% power except for brief to generator outage on 12/27 to correct-failed lug bolts 3/2/73 on A'phace of main transformer.
Brief power reductions also occurred for ucekly valve exercising.
During this period one' diesel-generator starting system malfunctioned i
due to a dirty air motor. All other air motors were inspected.
4 4
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4 Tabic III - Reactor' Scrams 7/10/72 Caused by turbine lockout initiated by loss of field relay.
Cause was a faulty connection in the generator amplidyne control circuit.
7/21/72
'n'ater scepage into control box caused spurious initintion of the deluge system for the main transformer.
The "A" phase transformer bushing failed, and the ratn transformer protective relay initiated a generator lock-out and scram.
S/26/73 Error in valving the Mechanical Pressure Regulator 1(MPR) into service caused high reactor-water IcVel condition which initiated scram.
6/16/73 Reactor scrammed from control valve fast closure signal caused by a worn turbine speed governor drive gear.
11/6/72 Accidental jarring of a steam flow transmitter during a routino instrument surveillance test caused scram and Group I isolation.
6/10/74 A hydrogen detonation in the recombiner system caused isolation of thi cir cjcctere. The reactor eu'u>equeatly
- scrammed on low condenser vacuum.
6/19/74 A generator lockout and reactor scram occurred due to a failed insulator on the 345 KV transmission line leading to the switchyard.
i l
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Table IV - Violatione A.
Violations identified during special Mana;;ement Inspection, May-June 1972:-
Description Corrective __ Actions 1.
Reactor was operated with 1-4. Surveillance test procedures loop A of the RilR service were revised to more clearly water system inoperative indicote acceptance requirements between June 1 and September and reasons for tests. The 21, 1971.
licensee initiated review of test _results by_the Shift 2.
The reactor was operated Supervisor and a designated between March 1 and the end-system engineer. A veckly status of September 1971 with one RHR report of surveillance testing Service water pump inoperabic.
was also initiated.
3.
Surveillance test procedure was deficient in that it did not require RHR service water pumps No. 13 and'14 to be individually tested as required.
4.
Redundant system components were not tested as' required between September 23 and October 9,1971, when RHR service water pump No. 12 was out of service.
5.
Temporary changes to operating
- 5. Improved procedures governing i
procedures had not been re-the review and issue of procedure viewed and approved, and the changes were provided. Technical l
Safety Audit Committee had not Specifications were revised at the I
reviewed recommendations of the
_ licensee's request to more clearly l
Operations Committee nor advised
-indicate review and approval i~
management of their recommen-requirements.
i dations.
t 6.
There was no evidence chat the
- 6. These deficiencies were corrected effectiveness'of the retraining by'the establishment of a formal program had been evaluated as'
. retraining' program as required 4
required, and all required by 10 CFR 55.
subjects were not included.
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7.
. The safety Audit Committee failed 7.
The licensce's response noted l
l to take the required action on an that the preventive. maintenance item of noncompliance brought to program had been under develop-i
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its attention, that being the lack cent since late 1970.
It was
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of a preventive maintenance program _
subsequently completed.
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f or instrumentation.
8.
Operations Committec-procedures 8.
Committoc' bylaws were~ revised
~
1acked specific instructions to include the required describing the content and method instructions.
of submit.sion of presentations to the Committee.
9.
Several deficiencies were obscr-9.
Procedures were modified cr.
ved with respect to the_ develop-issued in cach case to satisfy j '
ment, review, and implementation pertinent requirements.
of procedures. These involved:
(a) operation of drywell leak i
rate monitoring equipuent, (b) periodic review of -operating i
procedures, (c) procedure dis-tribution, (d) recording of test l
results, (e) filing of work request-authorizations, (f) revisions to procedures to reflect dnsign changes, (g) surveillance testing status report submission,- (h) review of Operations Committee minutes by the Safety Audit.
Committee and -(1) lack 'of operating procedures for abnormal leak rate.
\\
10.
Written procedures:had not.been 10.
Required procedures were issued i.
written or made available to all in August, 1972, i
station personnel for the-respiratory protection program.
j 11.
Test procedures for calibration 11.
The-licensee's.rcsponse noted the procedures to be available l
and preventive maintenance had in vendor's manuals.
Plant not been developed for installed' instruments used.to verify proper procedures were subsequently l
operation. of the. RiiR service-issued for specific instruments.
water system.
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Scheduled outage for refueling, turbine overhaul, 3/3/73 and miscellaneous plant nodifications, inspections, and j
r Items accomplished included:
(1) primary to 5/12/73 maintenance.
(2) containment local and integrated leak tests, installation of venturi in HPCI steam line, (3) replacement of MS1V operators and air accumulators, (4) modification of torus-drywell vacuum breakers, and On 3/30 No. 12 (5) modification of relief valves.
standby gas treatment systen calfunctioned due to improper Both SGT systems racking in of the fan motor breaker.
These were found to have improperly sized fuses.
conditions were corrected.
Fuel sipping indicated 25 These were fuel bundles to be potentially leaking.
An replaced with new or reconstituted fuel bundles.
injection valve in the LPCI system failed to open on 4/5 due to dirty contacts which were subsequently cicaned.
One starting system on No. 11 diesel generator All malfunctioned due to a dirty air line lubricator.
lubricators were cleaned and placed on the six-month inspection schedule.
Continued Established initial criticality for Cycle 2.
reactor heatup throughout the week and conducted HPCI, 5/13/73 Performed RCIC, and relief valve operability tests.
portions of main steam line transient test.
t HPCI turbine control system talfunctioned due to a 5/18/73 loose drive coupling, loose and worn gears, and a poor electrical connection. These problems were corrected and the UPCI system was returned to service within the allowed 7 days.
Gradually increased to 95% power by 5/19/73 Generator on line.
tests.
5/24. Completed special steam line transient The i
Scheduled shutdown to change control rod sequence.
5/25/73 sodium pentaborate tank was found to have'a low concentration and chemicals were added.
Resumed power operation. At 50% power a high reactor 5/26/73 water level scram occurred due to an error in valving the mechanical pressure regulator into service.
Resumed power operation at 90 to 100% except for weekly 5/27/73 valve exercising and.a reduction to 70% on 5/30 for to control rod withdrawal.
4 6/15/73 Reactor scram from turbine' control valve fast closure 6/16/73 The drive gears caused by worn governor' drive gear.
were replaced and improvements were made to prevent-recurrence.
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_4-2 6/20/73 Resumed operation at 100;; power except for brjef to reductions to withdraw control rods and conduct valve 7/30/73 exercising.
7/31/73 Scheduled shutdown to change control rod sequence, to inspect hydraulic nhock suppressors, and-perform-shutdown 8/9/73 margin tests.
Rebuilt 32 shock suppressors due to seal deterioration. ::SIV 80A operated slowly during a surveillance test. The yoke guide rollers were adjusted,
(
yoke guides were c3eaned and lubricated, and dashpot oil level was corrected to provide proper operation.
8/10/73 Resumed power operation.
During startup the outboard to main steam line drain valve could not be closed and an 9/7/73 interlock was adjusted.
Operated at 100% power except for load reductions to withdraw control rods, perfoto i
valve exercises, and repair a leaking feedvater pump instrument tap.
9/8/73 Reduced power and fully inserted control rod 14-27 to allow repair of a leaking scram exhaust valve.
Returned to 100% power 9/10/7'.
a 9/13/73 Discontinued control rod withdrawals to maintain 100%
to power due to end-of-cycle reactivity limitations.
9/23/73 Reactor power slowly decreased to 97%.-
9/24/73 Inserted control rods to establish rod pattern at 93%
power based upon revised reactivity calculations.
9/25/73 Reactor power further decreased to 92% while control l
rod inventory was held constant.
9/28/73 Scheduled shutdown to modify relief: valves, increase to safety valve setpoints, inspect hydraulic shock j
10/2/73 suppressors, and install connections to the modified off-gas system.
10/3/73 During functional testing of relief. valves-at low to reactor pressure, the "A" relief valve failed to close.
10/5/73 The air operator and second stage stems were found bent and corrected. Stems were also inspected-in other relief valves.
10/6/73 Returned to power operation,-performed operability to tests on all relief valves.
Increased to 92% of rated 10/8/73 power after repairing leaking feedwater pump instrument tap.
10/9/73 Reactor power-decreased slowly. as control rod inventory n
E to was held constant. AEC authorized relaxation of rod 10/17/73 inventory restriction on 10/18 based on revised transient I
analysis calculations.
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- 10/19/73 Increased to and operated at full power execpt for brief to reductions to withdraw control rods and perform valve 11/5/73 exercising.
l 11/6/73 Reactor scram due to inadvertent trip of main steam line high fJow censor during surveillance test.
Itesumed power operation.
11/7/73 Resumed power operation at 300% cxcept for a brief to reduction to reduce off-gac release rate below 103,000 11/13/73 uci/sec. During this period a reactor building-to-torus vacuun breaher did not operate freely. Shaft clearances were increased to provide proper operation.
11/14/73 Scheduled outage to rebuild hydraulde shock suppresrers, j
to install piping connecticns and perform preoperational 11/.17/73 testing of the off-gas holdup system, and repair a leaking feedwater check valve.
11/18/73 Resumed power operation at 100% except for brief to reductions to withdraw control rods and control off-gas 12/17/73 release rate.
12/8/73 Discontinued control rod movement due to end-of-cycle-to scram rei.ctivity considerations.
Reactor power slowly 1/13/74 decrenced to 917 on 1/1/74, remnicad at 41% until 1/13/74
[
1/14/74 Power operation continued with power -level reduced to i
to 83-90% to maintain off-gas release rate below 100,000 2/14/74 uCi/sec.
j 2/15/74 Reduced power to 60% for valve exercising.- Two MSIV's l
to failed to close on 2/16 during surveillance test. The 1
2/18/74 reactor was placed in hot standby with steam lines isolated while MSlV AC solenoid valves were modified and cleaned.
l l
2/19/74 Resumed power operation.
Reactor power was gradually to reduced'from 77% to'74% to maintain off-gas release-3/14/74 rate below 100,000 uCi/sec.
3/15/74 Scheduled octage to refuel the reactor and perform modi-to fications, maintenance, and testing.
Items accomplished 5/16/74 included:
(1) primary ' containment localLand integrated leak tests, (2) removal for the four originally installed safety. valves, (3) installation of four additional relief-valves with associated discharge piping to the torus, (4) i in-service inspection (5) installation of discharge spargers and vacuum breakers in HPCl and RClc turbine exhaust lines (to prevent water hammer and vibration during turbine operation), (6) modification of relief valves, (7) connection of the off-gas hold-up-system, and (8)
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,.,-s inspection of control rod blades for possibic inverted neutron absorber tubes. -Fuel nipping activities showed 83 fuel assemblica to be Icaking. These and others were replaced with 116 new 8 x 8 assceblics durina the outage, l
5/17/74 Establinhed initial criticality for Cycle 3.
Tested to llPCI, RCIC, and relief valves, and commenced operational 5/19/74 testing of the off-gas holdup system.
5/20/74 A hydrogen detonation occurred in the off-gas recombiner -
system, rupturing air ejector rupture discs.
Plant operation was resumed with the recombiner system bypassed.
5/21/74 Placed renerator on line and gradually increased power to to 97% of rated.
Power reductions occurred for control rod 6/5/74 scram timing tests and control rod withdrawal..During.
the period the HPCI auxiliary oil pump failed to operate because of a dislocated contact' assembly.
The contact s
was aligned and other similar contacts were inspected.
?
6/6/74 Sche.duled shutdown to return the off-gas holdup system to to service after modifying certain valves to eliminate 6/7/74 sparking.
6/8/74 Operated at low pcwcr for off-gas system testing.
to 6/9/74 6/10/74 A hydrogen detonation occurred in the recombiner system at 25%. reactor power.
Air ejector rupture discs ruptured. The reactor subsequently scrammed on low j'
condenser vacuum.
6/11/74 Resumed power operation with recombiner system-bypassed.
to
_ Increased to and operated at 93% rated power except 6/18/74
'for a brief, shutdown to repair a leak on a1feedwater 4
pump warmup line.
6/19/74-Reactor scrammed due to generator lockout caused by a l
failed insulator on the 345 KV: transmission line leading
_to the main switchyard.
Resumed power operation.
6/20/74 Continued operation at 92% of rated power. Reduced to to 88% on 6/23 to reduce off-gas release rate.
6/30/74 t
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I Table Ill - Reactor Scramn 7/10/72 Caused by turbine lockout initiated by loss of field i
relay. Cause uas a faulty connection in the generator amplidyne control circuit.
[
7/21/72 Water seepage into cont.rol box caused spurious initiation of the deluge system for the main transformer, 4
The "A" phase transformer bushing f ailed, and the rain j
1 transformer protective relay initiated a generator lock-out and scram.
s 5/26/73 Error in valving the Mechanical Pressure Regulator (:tPR) into service ccused high reactor water Icvel condition 4
which initiated scram.
l us 6/16/73 Reactor scrar ed from control valve fast closure sigr.al caused by a worn turbine speed governor drive gear.
11/6/72 Accidental jarring of a stena flow transmitter during a routine instrument surveillance test caused scram and i
Group I isolation.
6/10/74 A hydrogen detonation in the recombiner system caused 10cletien of thc cir cjcctaa. The teoucor du'udequently scranmed on lou condenser vacuum.
6/19/74 A generator lockout and reactor scram occurred due to a failed insulator on the 345 KV transmission line Icading to the switchyard.
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7.
The safety Audit Committee failed 7.
The licensee's response noted to take the required action on an that the preventive maintenance item of noncompliance brought to program had been under develop-its attention, that being the lack cent since late 1970.
It zas of a preventive r..afntenance program
-5.ubsequently completed.
for instrunentation.
8.
Operaticais Cornittee proceduren 8.
Comt.littee bylaws were revised lached specific instructions to include the required describing the content and method instructions, of submission of presentations to the Cormittee.
9.
Several deficiencies were obrer-9.
procedures vere mod'fied er i
ved with respect to the deve30p-issued in each case to saticfy ment, review, and implementation pertinent requirements.
of procedures.
These involved:
(a) operation of drywell leak rate conitoring equipacnt,- (b) periodic review of. operating procedures, (c) procedure dis-tribution, (d) reevrding of test results, (c) filing of work request authorizations, (f) revisions to rn refloce das1r,n procedurem changes, (g) surveillance testing status report cubnission, (h) review of Operations Committee l
minutes by the Safety Audit Committee and (1) lack of operating procedures for abnormal leak rate.
10.
Written procedures had not been 10.
Required procedures were issued written or made available to all in August, 1972.
station personnel for the respiratory protection program.
11.
Test procedures for calibration 11.
The licensee's response noted and preventive maintenance had the procedures to be available not been developed for installed in vendor's manuals, plant instruments used to verify proper procedures were subsequently operation of the RHR service issued for specific instruments, water system, l
1 l
l
p, 12.
A surveillance test and changes 12.
The surveillance test procedure to it had not been reviewed by was later determined to have the Operations Committee.
been approved, although the changes had not.
Revised in-structions for the writing, approval, and handling of surveillance procedures were issued.
13.
Certain operating and main-13.
Record storage techniques were tenance records were not kept improved to provide convenient in a manner convenient for record retrieval through cross-
- review, references and extensive use of microfilm techniques.
14.
Facility changes made prior 14.
These changes uere reported as to March 1971 had not been part of the semiannual operating-reported to the Commicsion.
report for July-December 1972.
15.
No formal quality assurance 15.
A program has been developed program had been implemented, and placed into use.
Regulatory operations review of the QA manual procedures was conducted in March 1974.
Inspection of program implementation and resolution of RO comments is pending.
B.
Other Violations:
Date and Description Corrective Actions 1.
- July 1972: Two of four drywell 1.
All drywell and torus pressure
. pressure sensing taps were sensing and sampling lines found to be covered with tape.
were inspected for-restrictions.
Signs were attached to. sensing taps-to warn against obstruction.
2.
- 0ctober 1972: Prompt action 2.
A loose connection in the low was not taken to restore stack flow annunciator circuit was sample flow after failure of a corrected.
A new surveillance sample pump.
test of the pumps and annun-ciator was established.
De-sign changes were subsequently made to increase the reliability of the sampling system.
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s 3.
- March 1973:
A torus manway 3.
Administrative instructions were cover was removed when primary issued to preclude recurrence, containment integrity was A warning was painted on both required.
manway covers requiring the Shift Supervisor's approval to remove them.
4.
- May 1973:
Sodium pentaborate 4.
procedures for additions to the solution in the standby liquid standby liquid control tank control tank uns inadvertently were modified to require (1) diluted to a concentration lecs campling before and after each than allowed by Technical addition and (2) written approval Specifications, of all additions by the Operations Supervisor and-Plant Chemist.
5.
May 1973:
Changes made to a 5.
The violation was-discussed with surveillance test procedure the individuals involved and a were approved by only one memo was issued to the technical licensed senior operator.
staff and licensed personnel reminding them of approval require-ments for temporary changes.
6.
July 1973:
The approved 6.
Local leak rate test procedures procedure was not followed were revised to more' clearly during an MSiv leak test.
indicate requirements.-
F. valuation showed the test results obtained to be valid..
l 7.
August 1973:
Fish, aquatic 7-8.
The licensee developed a vegetation, and other environ-laboratory quality control program mental samples were not on all
.and an administrative procedure occasions collected and analyzed for following sample progress.
l as required.
A consultant was to be retained to evaluate the overall program 8.
August 1973:
Records did not and develop audit procedures.
show several milk,. precipitation, Reinspection-of corrective action and vegetation samples to have is pending.
been analyzed as required.
9.
- 0ctober 1973: The stack' 9.
The local purge valve control effluent monitor was inadver-
' switch (which had been-lef t in tently rendered inoperable the " purge" position-to cause between August'7 and 10, 1973.
.the inoperability) was removed.
-This permits operation only'from the control room where a red indicator l
light. indicates that the monitor 1s in the purge mode.
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10.
- 0ctober 1973:
Surveillance 10.
The maintenance procedure was testing of ECCS components was revised to indicate the specific inadvertently omitted prior to surveillance requirements.
A removing a diesel nenerator mcmo was also issued to remind from service for maintenance.
shift Supervisors of the requirements.
11.
- December 1973: TFe volume of 11.
The violation resulted from an sodium pentaborate snjution indicator reference error, was less than the rc juired Additions ucre made to bring amount between August 28 and the tank volume within requirt-Septe.riber 4, 1973.
ments.
The indicator was recalibrated for the new reference.
Similar indicators were checked and found satisfactory.
12.
- December 1973:
Daily linear 12.
The Daily Log was modified to heat generation rate deterni-incorporate all daily test nations were not perforced on requirements.
November 9, 1973.
13.
January 1974:
Counting 13.
The 1dcensee's response stated frequency for reacter building that all cartridges would be vent filter cartridges was not counted daily until a clarifying it.creu=ed ou ually as required Technical Specifications chante when the groes beta-gamma release could be obtained.
Regulatory rate increased above 25% of the Operations review of corrective Technical Specifications annual actions is pending.
average limit.
14.
- January 1974: The licensee 14.
Procedures for receiving radio-i possessed approximately 3.4.
active materials were revised mil 11 curies of cobalt 60 on to prevent receipt of quantities June 6, 1973 which was not in excess of those allowed by authorized by the facility the license.
The licensee license.
also requested and obtained an amendment to the facility license authorizing possession of the cobalt 60.
15.
March 1974: The reactor-15-16.
All chemistry _ surveillance pro-l coolant was not analyzed for cedures required by Technical gross beta-gamma during the Specifications were revised to period February 14-21, 1974 include the purpose and require-(required every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />). A ments of each analysis. Memoranda; i
f monthly isotopic analynis of the were also issued to the coolant was not performed-during chemist ry group dot ining testing November 1973.
requirements.
A table included with the memoranda specified 16.
March 1974:
Reactor coolant which test s were required f o r chloride and conductivity each plant condition, analyses vere not performed every four hourn as required during the period l'cbruary 16-13, 197'., while operating at lou steam flow.
17.
March 1974:
Four inntances were 17.
Procedures were revised and/or noted of failure to follow personnel instructed as quality asrurance procedures neceasary to provent recurrence.
related to the installation of More frequent audits were additional relief valves.
The scheduled by the licensee.
instances involved (a) weld rod storage, (b) handling of non-conformance reports, (c) processing of purchase orders, and (d) receipt inspections.
18.
March 1974:
Planned audits had 18.
An audit was performed and all had not been perforned by the open items were resolved.
licenccc cr cnc cf hia ccatractors to deternine the effectivness of their portions of the Quality Assurance Program.
19.
March 1974: Audits were perforced 19.
Checklists were prepared and by a contractor without the use used during a subsequent audit, of checklists, and audit results Auditors were instructed in were not distributed as required the required distribution of by procedures, audit reports.
20.
May 1974:
The liquid effluent 20.
The licensee's response stated monitor was not being calibrated that the surveillance being monthly as required.
performed was believed to satisfy Technical Specifications.
- lndicates violations discovered and reported by the licensee.