IR 05000263/1974002
| ML20024G091 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/08/1974 |
| From: | Dance H, Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20024G089 | List: |
| References | |
| 50-263-74-02, 50-263-74-2, NUDOCS 9102070522 | |
| Download: ML20024G091 (17) | |
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U. S. ATOMIC ENERGY C0:0!ISSION DIRECTORATE OF PIGULATORY OPERATIONS
REGION III
Report of Operations Inspcetion RO Inspection Report No. 050-263/74-02 Licensec: Northctn States Power Company 414 Nicollet Hall Minneapolis, Minnesota 55401
)!onticello ::ucicar Generating Plant License No. DPR-22 Monticc310, Minnesota Category:
C Type of Licencee:
Eb'R (GE) 545 Mwe Type of Inspection:
Routine, Unannounced
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Dates of Increction: March 5 - 8, 1974 Dates of Previous Inspection:
February 27 - March 1, 1974 (Radiation Protection)
Principal Inspector:
P. !. Iu(nsonf t le f 7I (Date)
Accorpanying Inspector:
11. C. Dance Other Acccrpanyir.g Percennel: lione d.'l.. b,
,I/.(/ 7/
J o/ /..
!!. C Dauce, Senior Reactor Inspector Reviewed By:
'lk'R Operations (date)
9102070522 740409 PDR ADOCK 05000263 O
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StHMAltY OF FINDit;GS
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Enforcenent Action The following violationn are considered to be of Category 11 severity:
A.
Technical Specificaticn 4.6.C.1 requires that a reactor coolat.t sample be taken to determine (1) a gross beta activity at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, and (2) an isotopic analysis at least once per month.
Contrary to the above (Paragraph 6.a):
1.
The reactor coolant was not analyzed for gross beta activity between the period February 14 and 21, 1974.
2.
An isotopic analynis of the reactor coolant was nct performed during Noverber 1973.
B.
Technical Specificatien 4.6.C.2 requircu that during stent.ing rates below 100,000 pounds per hour, a sample of reactor coolant be analyzed every f our hours f or conductivity and chloride contcut.
Contrary to the above, such analytcs were not perf orr.c d during the interval of February 16 - 18, 1974, while steaming at less than 100,000
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pounde per hour. (Paragraph 6.a)
hicennee Action on Prcvious]v Identified I'nforcerent %:ttts The licennec has completed corrective actions related to items 5.a. 6 and 9 b as identified in the r,0:hQ cnierccr.cn Icttcr f clh inn the tiny 1972 canagement audit.
(Paragraphs 4 and 5)
Unusual Occurrencen A.
An F.CIC steam line high area terperature switch was found on January 29, 1974, to have drifted outside its allowed limiting setroint.
(Paragraph 12)
B.
The "A" RilR torus cooling injection valve operator totor failed on February 3, 1973, due to overheated noter vindings.
(Pcragraph 11.f)
C.
Tuo nain steam iso 3ation valves failed to close during a routine surveillance test on February 16, 1974 ("aragraph 9)
Other Sirnificint Findinte: None.
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Management Intervifw The inspectors conducted a canagement interview with Messrs. Neils (NSP General Superintendent, Nuc1 car Power Plcnt Operation), Larson (Plant Manager), and supervisory acmbur of the plant staf f at the conclu. ion of the inspection, The followinr. matter > were discussed:
A.
The unusual occurrences reviewed during the inspection and the licensee's related plans were briefly discussed.
(Paragraphs 9, 11.f and 12)
i B.
The inspector discussed his review of activities related to the of f-gas system, noting that he had no comments related to the conduct of the preoperationc1 testing program. He stated that further review would be given to planned retreatment of the A recombiner vessel and modification of recombiner heater control circuitry.
(Paragraph 14)
C.
The inspector stated that based upon review of the licensee's related corrective actions, violations fr9m the May 1972 managenent audit
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related to Operating Manual review and Volume P Memos were considered to have been corrected, but that a followup examination of these creas would be conducted in late 1974. The inspector also stated that in view of the licensec's retraining program that had been submitted to Licensing, the related violation from the same audit was also con-sidered to have been corrected.
(Paragraphs 4 and 5)
D.
The licensee'vas reminded to ensure that APP,M flow-biased scram set-
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poiato, after correction for APnN gain, remain within Technical Specification limits.
(Paragraph 7.c)
E.
The violations involving the omission of a reactor water isotopic analysis durfrq Novceber, and conductivity end chloridt requirements during lou secaming rates in February were identified. ihe licensee acknowledgcd the findings.
Subsequeatly, the inspector informed the licensee of the absence of the gross beta activity analysis during the same February period.
(Paragraph 6.a)
F.
The inspector stated that the Technical Specificationo listed the UPCI discharge pressure range as 150-1150 psig although Technical Specifica-tions Change Request No. 3 had requested a change to 1120 psig. This request has since been cancelled.
The licensee stated that a request for Technical Specification change would be tcsubmitted. The licensee stated that following test line modifications during the scheduled out-age, the HPCI discharge pressure would be deconstrabic over its full range.
(Paragraph 11.d)
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REPORT DETAILS 1.
Persons Contacted C. Larson, Plant Manager M. Clarity, Superintendent, Plant Engineering and Radiation Protection W. Andersen, Superintendent, Operation and Maintenance L. Eliason, Radiation Protection Engineer G. Jacobson, Plant Engineer Technical D. Antony, Plant Engineer, Operations S. Pearson, Stift Supervisor B. Day, Engineer F. Fey, Assistant Radiation Protection Engineer M. Hammer, Engineer J. Heneage, Engineer W. Hill, Engineer R. Jacobson, Plant Chemist B. Jenness, Engineer D. Nevinski, Enginect, Nuclear L. Nolan, Engineer J. Pasch, Engineer R. Perry, Engineer W. Shamla, Engineer, instruments
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2.
General The Monticello plcnt was operating at a reduced power level of 76% at the time of the inspection to maintain stack release rate below en administrative limit of 100,000 uCi/sec. The plant was scheduled to shut doun on March 14, 1974, for a refueling outaae of approxinately 11 weeks' duration.
3, Log and Records Revicu The following records were examined during the inspection without coraments a.
Reactor and Contror Room Lcg - February 16 - 20, 1974.
b.
Operations Conmittee Minutes - October 10, 1973 - January 16, and February 1, 14 and 15, 1974.
Safety Audit Committee Minutcu - November 9, 1973 and January 9 - 10, c.
1974.
d.
Weekly battery readings for No. 13 250 volt battery, October 30, 1973.
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t 4.
Actraining Program item 6 of the enforcement letter 1/ following the May 1972 manatement audit identified certain aspects of the retraining ptogram which did not comply with Technical Specifications requirements.
T?.a inspector noted during t.ie inspection that this violation had bec. ' /crecced by the formal retraining program which was submitted 2/ by b* licensec to the Directorate of Licensing in response to Appet Lh A to 10 CFR 55.
5.
Operatinn Procedures / Volume F Memog Item 9, Part b, of the enforcement 1ctter2./ f ollowing the May 1972 management audit noted that semiannual reviews of the operations manual had not been completed as required.
The liccnsce's response.1/ to the enforcement letter stated that a new review schedule had been established which would beconc effective following the first rewrite of each manual section.
Exanination of manual review reccrds by the inspector showed that 78 of the 100 manual sections had been revised and reviewed by the Operations Cornittee, with most of the remainder in progress. A repre-sentative stated that reviews of several of the sections of Volume A, General Administration, were being held in abeyance pending issue of the Administrative Controls Manual, which will supplant siCnificant portic:.s of the present Volume A.
A spot check of the status list maintained for manual revisions against Operationc Committee ninutes revceled no dir-(,,
crepancies.
Peric.dtc reviews subsequent to the initial review were noted to be proceeding on schedule.
A representative stated that an
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additional change to the re"few schedule had been cpproved Fy the Operations Cermittee, to the effect that the routine pericdic review 02 radiation saf ety procedures, Voltce E, had becn changes f rom annual to biennial, except for E.2 (Emergency Plan), which would continue to be icvicwed annually.
Item 5, Part a, or t's t
..orcement letter 5/ cited noncompliance related to temporary changes c+,
'ing procedures.
Review of the master copy of
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Volume F Menos hep in i. e control room shewed 84 to be in ef fect as compared to 206 in November 1972.
Some of those remaining in effect were to be deleted by pending revis1ons to the Operations Manual. A February 1974 revision to Sc? tion A.6, Plant Operating Practices, of the Opera -
tions Manual was noted to have provided more detailed Fuidelines for the 1/ Letter, RO:HQ to SSP, dated 10/19/72.
2/ Letter, ESP to DOL, dated 12/17/73.
3/ Letter, RO:HQ to NSP; datc; 10/19/72.
4/ Lettc, NSP tc RO:nQ, dated 11/10/72.
5/ Letter, RO:HQ to USP, dated 10/19/72.
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review, approval, and issuance of Volume F, Temporary Menos. L'hrea categories of memos--description, orders and procedurcs--are defined, the latter Ewo of which require Operations Committee approval within thirty days. Cancelled Volume F memos were noted to have been removed from the control room copy, although several inconsistencies in the index and the manual chapter cross-reference list were noted.
The incpector stated that baced upon 1:rprovements shova by the licensee in the review of operating procedures and Volume F memos and in vicu of the significant reduction in the number of Volume F memos in effect, the related violations were considered to be resolved, although a f ollowup review of the two areas was planned for late 1974, 6.
Coolant Chemictry Review of reactor coolant analyses between December 31, 1973 aad February 21, 1974 (except as noted) determined the fo13owing with respect to Technical Specification 4.5.C and 4.6.C:
(1) A reactor coolant sample was not taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> as required by Technical Specification 4.6.C.1(a) between February 14-21, 1974, and analy:cd for gross beta activity.
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(2) A reactor coolant sample was not taken every four houra se required by Technical Epecification 4.6.C.2 and analy?cd for conductivity and chloride contcut between February 16-18, 1974, while steaming at less than 100,000 pounds pcr hour.
(3) An isotepic analyrfs of the reactor coolant systm cac not per-formed during Novenber 1973 as rcquired by Technical Specific-ation 4. 6.C.1(b).
The surveillcnce data cheet on November 20, 1973 indicated an isotopic analysis (gamma scan) had been performed. Diccessions with plant perstnnci established that this data sheet is initiated when the gamma scan is initiated.
Due to a laboratory mix-up the saeple was not counted long enough and the work wan never completed.
The licensee's system of monitoring required surveillance testing had not detected the above ot.issions.
Review of analyses f rem October 1973 - February I?74, indicated typical values were as follows:
Total Iodine
3. 2 (Nov) - 1.65 (Feb) uCi/ml Chlorine
10 ppb Conductisity
0.14 - 0.7 umho/cm Gross Beta
1.3
- 2.2 uCi/ml-6-
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b.
Coolant Leakare Reactor coolant leakage rate was confirmed to be set up to complete on daily basis.
On February 18, 1974, the leakage calculation was confirmed to be within the limits of Technical Specifications 3.6.D and 4.6.D and as also indicated on sumnary data shocts for the period 12/1/73 to 1/31/74.
Review of the recently installed reactor aussel leak detection system showed it to be operating essentially an described in a
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licensec 1ctter.5/ Continuous recording and indication of floor drain and equipment drain sump icvels were noted to be available to the operator, plus a computer poir.t which permit readout of sump level and rate of change (in CTM) at any time, based upon computer inputs at 15-second interva3c. A licensee representative stated that some downward shift in the indicating range of the floor drain sump level indicator had been observed, with the effect that the indicator goes off scele low following pump-down of the surp. Refincuent of the indication was plcuned for the forthcoming refueling outage. The representative stated that system performance had otherwir.e been good, and that no difficult-1es had been experienced in th2 operation of the cump pumps cnd their controlling float switches. The floor drain leak rate in-dicated by the process cocputer, as confirard by observation of
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the floor drain sump pu=p run frequency, was noted to bc 0.09 gPm.
c.
Other Surveillance The recirculation system cross-tie interlock check required by Techulcr.1 Specification 4.5.1.1 was e c: freed to have been satis-factorily completed monthly from November 1973 threuth January 1974.
Reactor safety and relief valves were confirmed from a review of surveillance tests to have been tested and inspected as required by Technical Specification 4.6.E.
All safety valves vere set ct 1240 pcic and relief valves at i 1069 psig during the October 1973 outage as identified in RO Inspection I.cport No. 050-263/73-11.
Relief valve bellows leakage tests were confirmed to have been performed each thrce months betucen July 1973 and January 1974.
7.
Reactivity and Power Control a.
Control Rod Drives Revicu of CRD scrcn times during the current cycle er.tablished that 49 CRD's remain t.o be tested to meet Technical Specification 4.3.C.
Completion of this testing was confirmed to be scheduled immediately 6/ Letter,1:SP to DOL. dated 12/28/72.
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after shutdown for the refueling outage scheduled March 14, 1974. All data reviewed were taken from the multipoint recorder which connects 28 CRD's.
All scram times were within times contained in Technical Specification 3.3.C.
On November 6 the maximum 90% scram insertion time was 2.84 seconds.
Correlation between the multipoint recorder trace and individual rod insertion traces from the brush recorder was established.
Included in this review was the testing performed as a result of GE question-ing on September 28, 1973, the f ast CRD insert ion times f rom full out to 5% insertion.
Subsequent testing by NSP determined that 85 msec should be added to the scram times.
This interval was the demonstrated time delay that it takes pen No. 30 of the multipoint recorder to buildup to a printing threshold of 1.6S VDC.
Data since October 1973 were stated to be corrected by adding 85 msec.
Review of the November 6, 1973, was inconclusive to the inspector since the starting point was not clear.
Even considering tl.e above correction the CRD's meet the required specifications.
CRD stall flow testing during January and February 1974 indicated two drives with greater than 5 gpn.
Approximately 25 CRD's were scheduled to be replaced during the outage on the basis of past tests and routine change out.
Two drives scheduled for replacement are 18-31 and 22-35 which have been in service since startup and are the only two modified ([4 drives with the inner screen mounted on the stop piston, No plans exist for modifying other CRD's since scram times continue to be satisfactory.
Weekly control rod exercise tests required by Technical Specification 4.3.A.2 were reviewed for the period December 29, 1973 - February 23, 1974, and confirmed to have been completed.
Status of control room accumulator level and pressure alarms was con-firmed to be included in the Daily Surveillance Leg.
Complation was verified
- December 18, 1974.
b.
Intermediate Range Monitors (1!U1's)
The functional test of the SpJt rod block and the IPJi scram and rod block were confirmed to have been performed as required by Technical Specification Tabic 4.1.1. on February 16, 1974, prior to the reduction to low power operation.
The lpl1 requirements were confirmed for the outage bounded by the November 13 and November it, 1973 test.
Dis-cussions with instrument personnci confirmed that the deviation values in parenthesis on IpJi test No. 0013/0043 were be bg used as permissible drifts and not as "as left" settings.
Instrument Depar tment as lef t-8-
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m records indicated the IRM values of Technical Specification Table 3.1.1 were satisf actory f rom July 30, 1973 to February 16, 1974.
The IRM heat balance calibration was confirmed to have been completed for the shutdown and subsequent startups beginning Noveciber 14, 1973 and February 16, 1974. The calibration appears difficult with questionable accuracy at the low power with changing conditions. Scttings were considered conservative.
Date Heat Balance %
IRM APRM February 18, 1974 0.88
3 Noven.ber 18, 1973 1.87 2.3-7
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c.
Averanc Power Ranne Monitors (APP?1'sl APRM heat balance calibrations perforned were reviewed f or the interval January 2 - February 25, 1974 and found satisfactory.
In gerneral APRM output signals were being Icf t 0.5 - 1.5% higher then calculated thermal pover.
All APRM channels were noted to be con-servatively set on March 5, 1974.
Cetputer heat balances compared within 0.5% to canual heat balances for the review period February 1 - 15 and March 1 - 4, 1974.
The calculatiens are confirmed three (.)
times per week.
g, APLM weckly functional scram tests for the period : ove:rber 1973 -
January 1974 required by Technical Specificatica Tabic 4.1.1 were reviewed and found satisfactory.
Calibration data for APP.'! flow-biared ceram and rod bloch were reviewed against Technical Specifications requirements, and were note t to comply when the conscrvative APRM gain settings used by the licensco were taken into account. The APRM flow-biased scram setpoint corre;ponding to 50% rceirculation driving fic.. vas noted to have been Icft at 87.6-87.9% for all channels on January 24, 1974, but APRM indicated power was noted to bc generally 0.5 to 1.5% con-servative during this period. Yh: l'icensee was reminded to ensure that the trip points, after adjustecnt f or APRM gain, remain within Technical Specifications limits.
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d.
Core Checks l-l Reactivity anomaly checks were confirmed to have been perf ormed conthly f rem June 1973 through 'iarch 1974 by comparing rod inv ertion values to GL provided curves.
Core reactivity has continued to decrease as predicted since the cycle 2 st artup.
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The peak heat flux determination was confitcod to be setup routinely on a daily basis as required by Technical Specification 4.1.B.
On February 1,1974, at 82% power the peak heat flux was calculated to be 289,000 BTU /hr-ft2 with a peaking factor of 2.7.
The latter was determined by using conservative type curves provided operating personnel.
Computer calculation indicated the peaking factor to be 2.2.
c.
Other Surveillance The Main Steam Line Isolation Yalve Closure Scram Test Procedure (Test No. 0008) was reviewed to assure that cach of the contacts and relays are tested. The test required by Technical Specifica-tion 4.1.1 was found satisf actory.
Reactor liigh Pressure scram settings were found to have been set properly between November 1973 and February 1974 as required by Technical Specification 3.1.A and 4.1.A.
The bases for the pro-cedural retpoint value was confirmed by tracing systen elevations.
f.
no Bulletin No. 73-6 The above bulletin dated November 27, 1973 requested licensees to describe their administrative control system of coordinating core movements to prevcnt an inadvertent criticality. The 31cenace's (44 response dated January 10, 1974, was determined to be, satisfactory based on previous reviews of procedural and administrative controls.
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As stated in the responic, the liccnsee verbally indiccted that improvements in coordinating core tuncuvers were being reviewed.
8.
Refueling Preparations A facility rcpresentative stated during discussions with the inspector that the new fuel to bc inserted into the core during the refueling outage had been inspected and placed in the new fuel storage vault. He explained to the inspector new procedurcs which will use the assistance of a coeputer to prepare the sequence used in the moving of fuel and other core components during the outege.
The inspector reviewed reports of inspections of the reactor building cranc and refueling pictform conducted in Octobcr 1972, by a crane I
inspector from an outside firm. The reports indicated no unsatisfac-tory conditions. A facility representative stated that a similar inspection by the same firm had bcer. ccnducted during the lart week of February 1974, and that a report was expected in the near future.
The inspector excmined procedurcs fer semiannual inspection of the reactor building crane which were being sent to the operations l
Committee for review prior to issue. Requirements for magnetic part-
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icle testing of the main hoist hook and dye penetrant test of the
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auxiliary hoist hook were to be added.
The inspector questioned the licensee's fulfillment of the commitment for semiannual crane inspection expressed in Section 10.2 of the FSAR. The licensee representative referred to the inspections by the outside firm prior to the lif ting of heavy loads during the previous and the forthcoming refueling outages and stated that issue of the new procedures would provide for semi-annual inspections thereafter.
9.
Failure of Two Main Steam Isclation Valves (MSIV's) to Close A licensee report 1/ discussed the failure of the outboard MS1V's in the B and C main steam linec to clore during a routine surveillance test on February 16, 1974. The report also discussed corrective actions taken, including repairs perfort:cd on the air solenoid valves of all 8 MSIV's.
Discussions wit h licensec representatives and review of a schematic diagram showed that (1) e differential pressure (equal to instrument air pressure) is normally applied across the viton seat of the AC solenoid, (2) air pressure ic norm-ally applied to both sides of the DC solenoid valve, such that no differential pressure existr, (3) seat deformation on the DC solenoid was not significant, (4) based on recornendations frou the vendor, spring-loaded sects ucre net installed in the DC solenoid valves, and (5) the metal chips discusced in the licensec's rcycrt were not consid-ered to be a significant fccinr in the na1 functions observed.
Surveil-lance test records showed that all MS1V's had closed in the required ("~"
3-5 seconds during tests folloving the repairs. The inspector asked whether spring cuchiened up; r ccats m3glt eventually be necessary on the DC colenoids, but otherwise had no comments on the corrective actions taken by the licensee.
A liccunee representctive stated that overhaul of all solenoid valves under the supervision of a vendor representative war planned during the forthcoming refueling outage, and that further evaluation of calenoid vclve performance would be made at that time.
10. Vane Type Flow Evitchen A licensce representative stcted during the inspection that the vane type flow switch installed in the standby liquid centrol system was to be removed during the forthcoming refueling outage, in that it is not required for proper system operation, lie stated that the stubs of the original flow suitch paddles were still opcrating satisf actorily in the residual heat removal system, although further improvements were being planncd f or this system and f or the reacter water cicanup system. These would likely utilize annubar-type (pitotstatic) indica-tors, and would probably not be installed during the 1974 refuelink outage.
2/ Letter, NSP to DOL, dated 2/25/74.
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11. Emergency Core Cooling System Review established that test procedures, most of which have been rewritten in an improved format, are provided for each of the required surveillance tests. Bases for many of the procedural setpoints were not readily availabic at the station or in the procedure; for instance, the core spray pump discharge pressure equivalent to a reactor discharge pressure specified in the Technical Specifications. The latter was determined to be satisf actory f rom discussion with plant personnel and review of test results prior to initial reactor startup. Other systems values appeared reasonable but were not rechecked.
The RHR sub-system and HPCI were confirmed tu be properly set-up on the control room pancis during the inspection.
Equipment in the RHR sub-system pump rooms was noted to be satisfactorily lined up, s.
Low Pressure Coolant Injection (LPCI)
Testing requirements contained in Technical Specification 4.5.B.1 were confirmed to have been performed for the intervals noted: quarterly flow rate, Nover..ber 1973 - February 1974; pump operability, June 1973 -
February 1974; and MOV operability, October 1973 - February if L.
CV-1995, No. 12 pump minitum flcu valve, closed automatically on low flow during testing on January 2,1974. From the work request form fy and discussion with plant personnel it was established that the low 4 *'
flow indicating switch uns loose, resulting in a enlibration shift.
The valve remained operable in the manual mode from the control roct.
Flow testing was satisf acterily pcrfort.ed f ollorir.c r.intenance.
b.
Core Sprov Svr.t ens Testing requirements in Technical Specification 4.5. A.1 were confirmed to have been succersfully completed for periods shown:
quarterly flow rate:
January - February 1974 (three tects); nonthly pump op.ra-bility: October 1973 - February 1974; monthly UOV operability: July 1973 - February 1974; monthly header t.P and calibration: October 19 73 -
January 1974 (four tests); and daily header :.P chech: October 31, 1973 and February 1, 1974.
c.
Residual Heat T.cmoval Service Water Syntem Quarter]y f]ow rate test requirements contained in Technical Speci-fication 4.5.C were found to have been satisfactorily performed for the June 1973 - February 1974 interval reviewed. Tests were noted to have been perforr ed on the "A" system fol]owing taintenance (Septerber 1973) and for a HPCI inspection ( June 1973).
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d.
High Pressure Coolant Injection System Surveillance tests designated in Technical Specification 4.5.D.1 were found to have been satisf actorily performed for the intervals reviewed as shown:
monthly pump operability:
September 1973 - February 2,1974; monthly MOV operability:
January - February, 1974; and quarterly flow rate: July 1973 - January 1974.
Difficulty has been experienced in simulating the required reactor pressure range during tests (typically values of 150 to 1120 are obtained) due to the oversized throttling valve (MO-2011) in the test line. The inspector reviewed a modification scheduled for the March 1974 outage to install a 6" self drag type valve to correct the difficulty. The pressure range of 150 - 1150 psig called for in the Technical specificatlons was identified in Change Request No. 3 (dated August 20, 1971) as in error and should read 150 - 1120 psig according to the licenece.
The licensec agreed to re-initiate action to revise the pressure range in the Technical Specifications.
Fol-lowing the above modification, the full designed flow and pressure range is expected to be demonstrated during each test.
The HPCI system's primary flow and pressure instrumentation, including system trips was noted to have been calibrated and cheched on January 28, (,,
1974, in accordar.cc with test procedure No.,7130.
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e.
ECCS Instrurentation Calibration of ECCS instranentation listed in Technical Specification Tabic 4.2.1 was confirmed to have been conducted as required for the period Novceter 1973 through February 3974.
Once per cycle tests were conducted in May 1973.
L Micellaneous core spray ficw and pressure instruments were confirmed l
to have been routinely checked each six tenths since January 1973.
f.
Torus Coolinc Injection Valve I
A licensec reportE/ described the failure of the "A" Torus Cooling reportgyn Valve, MO-2008 on February 3,1974.
inject A previous ifcensee discussed a similar failure of the same valve, although the inspector concluded the causes of the two failures to be unrelated.
(During preparaticn of the current inspection report, it was noted that the earlier licensee report had been incorrectly referenced in a i
f/ Letter, 11SP to DOL, dated 2/12/74.
9/ Letter, NSP to DOL, dated 11/7/73.
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previous inspection report 12/). Review of the occurrence and a discussion with licensee representatives showed the events to have been as described in the licensec's report.
The abnormal occurrence file contained a listing of ECCS motor operated valves which included 12 valves similar to MO-2008.
Examination of one of these valves by the inspector during a plant tour showed the stem clamp to have been tightened and staked as indicated in the licensce's report. The licensee representative stated that a review of system diagrams indicated that no similar valves were located in the drywc11, but that a review inside the drywell would be made during the refueling cutage, 12. RCIC Stean Line Terporature Switch A licensec report l/ discussed a condition wherein a steam line high l
area temperature switch associated with the reactor core isolation cooling ( RCIC) system tripped at a temperature greater than that allowed by Technical Specifications.
A similar occurrence (related to the high pressure coolant injection system) reported by the licensee was reviewed during a prcvious inspection.12/
It was noted during the earlier inspection that discuselons between the licensee and the manuf acturer were ongoint; with relation to the drif t experienced with tcuperature switches of thin type.
The licencec had also indicated in discussions with the inspector that rcre frequent calibration would not likely improve the perforucnce of the r."itchec, since they must bc (;q-removed for calibration and the increased handling would likely offset the gains of increased calibrction frequency.
A licenace representative stated during the cuncut inspectica that consideration was being given to new tcuperature menitors, probably using thcrtocouples, uhich could be calibrated in place and which gave promise of more reliabic operation.
The representative stated that the switch which most recently malfunc-tioned had been toroved from scrvice and that the pre.ctice of setting the tenperature switches at 10-150F below their Technical Specifications limit would continue pending evaluation of an alternate installation.
The ine.pector verified by review of calibration records that the as-found and as-1..f t actpoints of all other Ler..perature switheca usdo-ciated with tne RCIC cystem had been within Technical Specifications limits during calibraticns perforned on October 30, 1973 and Janua*y 29, 1974.
13. llydaulic Shock Supprer: sors and Est raint n
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A previour, inspection report 13./ discussed ncticnc taken by the licent.ce in response to Regulatory Operations Bulletin Mos. 73-3 and 73-4 and a 10/ RO Inspection Rpt No. 050-263/73-12.
11_/ Letter, MSP to 1>0L, dated 1/30/74.
12/ RO Inspection Rpt No. 050-263/73-11.
13] RO Inspection Ept No. 050-263/73-12.
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.i related letter from the Directorate of Licensing. A subsequent letterli/
from the licensee described a reinspection of all suppressors within the drywell on February 17, 1974 An attached table also described repairs performed to suppressors located outside primary containment. The inspector examined surveillance documentation of February 17, in-con-tainment suppressor inspection and of an inspection conducted on February 13 of suppressor units located outside of the dryvell. No exceptions to the conditions reported in the licensco's 3ctter were noted by the inspector. A licensco representative stated that suppressor units outside primary containment would be inspected once more prior to the refueling outare, and that a followup report was intended. lie stated that all supprecsor units had been reuorked using internal coft parts made of ethylene propylene and new oil-fill fittings having buna-H reats.
14. Off-Gas System A review of off-cas system testing during the inspection showed that all preoperational tests had been completed.
Operations Cotr.ittee Minutca revicued by the inspector indicated that all but four of the preoperational te ts had been revicwed by the Operations Cortmittee. A IJconsee representa-tive stated that three of these four had been reviewed in recent meetings for which minutes were not yet issued, and that a final tcct report for the completed T3-1C ventialatf en cynten tie-in van noen to be received from the test engineer. The inspector reviewed several of the completed tests
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for general content-and test results. The contents of each tes package
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were noted to be as described in c previnus inepection reportE The inspector made no comment on the portions of the systcu test pregram conducted to date.
Licensee representatives stated that plans called for tie-in of the off-gas system to the plant during the refueling outate, following corrective actions related to the recombiner and its heater control circuitry. Oper-ational testing of the of f-gas system would then be accomplished follow-ing plant startup after the cutage.
A corporate representative stated during a telephone discussion that metallurgical tests of the recombiner vessel material had been conpleted, leadinp to a conclusion that the vessel could be annealed in place to provide satisf actory performance.
lle stated that some reduction in yic]d strength of the lower portion of the vessel had occurred, but that the overthicPnens in the initial design would still provide adequate strength.
The vessel was also to be hydro-statically tested to its initial test pressure following the in-place anncaling process. The reprc.centative also stated that the henter con-trol circuits had been redesigned and would be codifjed accordingly 14/ Letter, NSP to DOL, dated 2/15/74.
15/ RO Inspection Rpt No. 050-263/73-08.
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"A during the outage. The inspector deret ted comment on plans for the
recombiner vessel and its heater controir, pending review of the licensco's evaluation of the proposed corrective actions.
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UNITED STATCS
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ATOMIC CNCRGY COMMISSION
DiferCTORATC Or REGULATORY OPERATIONS R E.Glotv Ill
- *iTI 799 ROOSCVELT ROAD 1rLEPefnNC GLEN E LLYN ILLINOIS 00137 (312)Et,0 20 0 Y
A.
R0 Inspection Report No.
050-263/74-02 Transmittal Date
April 9, 1974 Distribution:
Distribution:
B0 Chief FSLEB R0 Chief, FS6EB RO:HQ (S)
RO:HQ (4)
M Central Files L:D/D for Fuels & Materials Regulatory Standards (3)
DR Central Files Licensing (13)
R0 Files RO Files B.
RO Inquiry Report No.
Transmittal Date
Distribution:
Distribution:
RO Chief FSLEB R0 Chief, FSLEB RO:HQ (5)
RO:HQ DR Central Files DR Central Files Regulatory Standards (3)
RO Files Licensing (13)
RO Files (
C.
Incident Notification From:
(Licensee 6 Docket No. (or License No.)
t
Transmittal Date
l l
Distribution:
Distribution:
l RO Chief, FS6EB R0 Chief, FSLEB
!
RO:l'Q (4)
RO:NQ (4)
Licensing (4)
L:D/D f or Fuels 6 !bterials DR Central Files DR Central Files i
R0 Files RO Files l
l
!