IR 05000155/1976004
| ML20002D724 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/11/1976 |
| From: | Erb C, Hunter D, Jordon E, Kohler J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20002D723 | List: |
| References | |
| 50-155-76-04, 50-155-76-4, NUDOCS 8101220238 | |
| Download: ML20002D724 (19) | |
Text
{{#Wiki_filter:. . C. S. NUCLEAR RTfULATORY C0tef1SSION
' OFFICE OF INSPECTI!N AND ENTORCDfENT
- . REGION III-
g Report of Operations Inspection . . . IE Inspection Report No. 050-155/76-04 . Licensee: Consuners Power Company 212 West Michigan Avenue Jackson, Michigan 49210 Big Rock Point License No. DPR-6 . Charlevoix, Michigan Category: C Type of Licensec: BWR (GE), 240 !Ce ' Type of Inspection: Routine, Announced Dates of Inspection: February 2-6, 1976 ocA.59 i - Principal Inspector: D. R. Hunter 3/8/7b (Date) Accompanying Inspectors: J. E. Kohler , C. M. Erb . Other Acco:::panyin sonnel: None Reviewed By: J rd :i bf 3 // 76 . React Proj ects '(Ddte) Section No. 2 .
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. SUMMARY OF FINDINGS , i . Inspection Summary . ,- . Inspection of February 2-6, (76-04): Review of operations, refueling operations, reportabic occurrences, item of' noncompliance, plant cleanliness, review and audits, quality assurance, facility tour, and inspector identified and outstanding items.
- Enforcement Items None.
Licensee Action on Previously Identified Enforcement Items A review of plant modification controls indicates that the licensee actions are not completed.
(Paragraph Sg, Report Details III) Other Sinnificant Findings A.
Systems and Components Unresolved Item - The inspection results,. (radiography and ultrasonic tests) for four welds in the core spray line were lacking third party inspection. This item will be resolved prior to startup.
(Paragraph 2, teport Details II) B.
Facility Items (Plans and Procedures) 1.
The plant commenced shutdown on January 30, 1976, for.the refueling and major modification outage. The outage is scheduled to extend until about tby 15, 1976.
2.
The licensee had not reviewed the safety aspects of plant personnel regarding fuel handling activities in conjunction with an unlimited number of people inside the containment sphere. A preliminary offsite review was performed during the inspection which indicated no unacceptable personnel risk.
(Paragraph 8, Report Details I) C.
Managerial Items None.
D.
Noncompliance Identified and Corrected by Licensee None.
I 2- - . D s .r . .
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E.
Deviction3
- s none.
F.
Status of Previously Reported Unresolved Items , .
None.
. . Management Interviev ' The management interviews were conducted on February 5 and 6, 1976, by Messrs. Kohler, Erb, and Hunter with the following persons present: C. J. Hartman, Plant Superintendent R. B. DeWitt, Manager of Production - Nuclear R. B. Sewell, Nuclear Licensing Administrator C. E. Axtell, Chemistry and Radiation Protection Supervisor R.-W. Voll, Reactor Engineer D. E. DeMoor, Technical Superintendent C. R. Abel, Operations Superintendent - A. C. Sevener, Operating Supervisor R. E. Schrader, Instrument and Control Supervisor G. C. Tyson, Maintenance Superintendent S. E. Martin, Engineer G.-B. Szczotka, Quality Assurance Superintendent A.
The inspector stated that the relatively large number of people in ' the containment during. fueling activities appeared to warrant a review by the plant to determine the adequacy of the' containment exits.
The licensee acknowledged the statement and indicated that an offsite review had been performed based on the discussion with the inspectors and no unacceptable personnel hazards were revealed. The inspec-tor stated that the item, containment sphere evacuation, would be reviewed in a subsequent inspection.
(Paragraph 8, Report. Details I) . B.
The inspector stated that housekeeping in the containment sphere ~ during refueling was less than desirable particularly with respect to contractor installation of the Reactor Depressurization System.
The licensee stated that additional personnel were assigned during the outage and the area of plant housekeeping would be reviewed with the construction forces.
(Paragraph 5, Report Details I and Para-graph 6a, Report Details III) C.
The inspector stated that the step ladder and hose connection used to rig the transfer cask with an emergency supply of cooling water in the event of loss of offsite power were not designated as rafety equipment and were uncontrolled. The licensee stated-3-(' , . .I .
- - _ _ _ _ _ ___ _ ___ ____ __ ____ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _____-__ __ ___ ___ _. _ _ _. , ~ th:t caplo time cxicted ti rig th2 trcnsfdr ecck with cmergency , water and therefore these itcms need not be controlled. The inspector has no further questions retarding this item at this . time.
(Paragraph 9, Report Details I) D.
The inspector stated that review of the' licensee's a'ctions to correct containment instrumentation problems revealed no dis-crepancies.
(Paragraph 2, Report Details III) E.
The inspector stated that the review of operations revealed one item concerning the freezing of the stack gas monitor system for short periods of time during plant operation. The licensee stated that the freezing of the stack gas monitor systems is under review to determine corrective actions.
The licensee also stated that the outage times and air ejector off-gas activity were considered in the calculations of the radio-active releases from the site. The inspector stated that the stack gas monitor frcezing was not a new problem at the plant, and the plant should resolve this item expeditiously. The licensee acknculedged the statements made by the inspector.
(Paragraph 3d, Report Details III) F.
The inspector stated that a review of refueling operations revealed two ite=s concerning the refueling procedure.(RE-02) prerequisiter and the testing of the fuel transfer cask safety brake mechanism.
The licensee stated that the two areas will be reviewed and appro-priate actions taken.
(Paragraph 7a, Report Details I and Paragraph 4, Report Details III) G.
The inspector stated that a review of the item concerning the station battery supports will be reviewed further by IE:III and licensing.
The licensee acknowledged the stateuent.
(Paragraph Sa,. Report Details III) H.
The inspector stated that during the plant tour while exiting the a containment vessel personnel hatch, the operating =echanism failed to function properly. The inspector stated that the problem was identified and appropriate corrective actions planned.
The licensee acknowledged the statements.
(Paragraph 6c, Report Details III) I.
The inspector stated that the failure of the off-gas isolation system to effectively isolate the off-gas stream in the test performed immediately prior to the plant shutdown and the fact that the off-gas isolation system has failed previous tests,
- indicate a need for an engineering review to determine the unidentified flow path to the stack. The licensee stated that j
a further review would be provided.
(Paragraph 7, Report l Details Ill) -; -4- - . , f
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The inspector discussed the apparent need of an alternate path for- ' vacuum relief on the containment. The licensee stated that a-Consumers Power engineering review was in progress to determine-if an alternate path was actually required, and stated that the . - appropriate actions would be taken. -(Paragraph Sc. Report Details III) K.
The inspector stated that during the inspection of the core spray system welds performed in the 1975 outage, ques,tionable practices on specific welds were revealed.
1.
The inspector stated that he had reservations about'the quality of two welds made during the elevation of the four-inch core spray lines. The authorized inspector had not signed off on the radiography of four welds identified as North, South, IA and 14A.
(Paragraph 2c, Report Details II) 2.
The inspector noted that an ultrasonic baseline inspection had been madeoon the above four welds, with no indication of recordable back reflections from the veld root area, which is extremely irregular as shown by radiograph.. (Paragraph 2c, - Report Details II) L.
The inspecor stated that an examination of QA dccumentation for the RDS installation by Catalytic indicated that the work was being processed and accepted to applicable ASME Codes.
However, the QA audit by CP did not appear to bear acceptance signatures by CP. indicating their approval of material and welds being incorporated into the RDS.
(Paragraph 3, Report Details II) M.
The inspector noted the use of cicar poly around and near the spent fuci pool and the reactor vessel opening was considered questionable cone.idering the difficulty in locating poly under water.
The inspector asked the licensee to review the use of such materials.
The licensee stated that the question of using clear poly would be reviewed.
(Paragraph Sb, Report Details III) ! -5- . T -
RFPORT DETAILS .. , b Prepared By: - . . J. E. Kol er (Date) Reviewed By: / f [ [~. W.
S'. Little ADite) ~ . 1.
Persons Contacted C. J. Hartman, Plant Superintendent R. Voll, Reactor Engineer J. L. F.eumin, Maintenance Engineer C. E. Axtell, Chemistry and Radiction Protection Supervisor 2.
Pre-Refueling Activities The inspector verified that surveillance testing involving the pre-refueling activities have been completed.
a.
Preparation of the transfer cask for refueling (procedure MFIIS-1), b.
_ Crane testing - testing of the fuel handling cables (procedure MFilS-2).
c.
Refueling interlock test (procedure TR-02).
d.
Communication systems verification.
c.
Cooling capability for stored fuel.
3.
Refueling Deck Radiation Monitors Radiation monitoring on the refueling deck availabic for protection ' of the refueling crew consisted of a Continuous Air Monitor (CAM).
The monitor was checked for operability daily on_ day shift by verifying that there is flow through the uionitor and that the strip chart was functioning properly.
4.
Fuci Handling Activities The inspector verified by record review and direct observation that the following conditions existed and fuel handling activitics during refueling were being conducted according to approved proce-dures, a.
Core monitoring consisting of two source range monitors were reading greater than two counts per second.
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_ -, - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ ------ . - - _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - _ _ _ _ _ _. -. - - - - - -. _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ b.
Contcinment irtogrity tes maintcined-conricting cf the doubla - door cir lockn being clacrd and th] contcinment cphera vent ' valves open during fuel movement.
- . c.
The inspector verified that insertion and removal of fuel bundles was in accordance with approved fuel bundle removal procedure RE-02.
.J . , d.
The inspector verified that fuel accountability, consisting of a bundle unloading sequence specifying, spent fuel pit - locations, as well as tagging procedures in the control room, was in accordance with approved procedures.
Core' internals were protected with polyethylene.
e.
f.
The make-up of the refueling crew on the deck and in the control room was in.accordance with the established plant procedures. The minimum crew requirement consisted of a roving shift supervisor, one licensed operator in the control room, two licensed operators on the reactor deck and an auxi-liary crane operator.
g.
Water level and water temperature in both the spent fuel pit and the reactor were being monitored by control room operators and auxiliary plant operators.
h.
The re' actor mode switch _was in the refueling mode.
i.
The inspector verified that a licensed operator.was pre;ent in the control room and in constant communication with the fuel handling crew during all fuel movement.
' $. Housekeeping , llousekeeping inside the containment sphere-during fuel movement was Poor, particularly with respect to contractor personnel inside the reactor sphere.
The inspector noted unsecured power cables, welding bottles and litter in the sphere.
6.
Previously Reported Unresolved Item The inspector reviewed the unresolved item / pertaining to the fire I stop penetration sealant material qualification'. The fire stop modifications, indicating the uarecolved item, is still in engineer-ing and the fire stop field activitics have not commenced. This item remains open.
. 1/ IE:III Inspection Report No. 050-155/75-16.
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Previou]1y Reported Open Items , , ' -a.
Revisions to Refueling Procedura RE-02.- 1.
The inspector determined that fuel handling procedure RE-02 revisions included a bundle unloading sequence specifying spent ~ fuel pit location, and-the prerequisite section signoffpriortoanyfuelmovement.goceduresrequiring.
included the appropriate refueling-2.
The fuel handling procedure (RE-02) prerequisites did not contain all the required systems and components required by the Technical Specifications during refueling operations.
It was. determined through procedure review and discussions-with the licensee that the refueling procedure (RE-02) should cover the plant conditions and requirements = since the master checklist is not effective 'during this period ~ of time after plant shutdown until plant refueling..The inspector reviewed selected-syste=s required during the refueling operations by the Technical Specifications and noted no discrepancies.
b.
Ventilation Requirenents in l'uel Storage Areas The licensee is handling fuel with the containment vent valves open. Upon high radiation signal, the refueling crev vill notify the control roo= operator by telephone to isolate the containment sphere. NRR is aware of this procedure and is currently reviewing the necessit7 of automatic containment isolation on high radiation for in.sta11ation prior to the next refueling outage cycle 14). The inspector considers this item closed out.3 c.
Refueling Radiation Deck Mon Mrs The CAMS do not annunciate in control room, however, two area monitors required for monitoring criticality annunciate in the control room. The licensee stated that any significant fuel handling. incident would cause the control' room area monitors to annunciate. As stated above, subsequent control roos operator action vould be to close the containment sphere venti - lation isola * ion valves. The inspector considers this ite= closedout.N 2/ IE:III Inspection Report No. 050-155/76-01.
3/ Ibid.
, 4/ Ibid.
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Fuel Sipping
. . The procedure for dry sipping has not been completed by.the.
~ licensee. This item remains open and will be followed at the , next-inspection.
r.
Fuel Inspection . ., ' The procedure for-vendor fuel inspection an'd bundle recon-stitution of-reactor fuel has not been approved by the licensee.
This item remains open and will be followed at the'next inspec-tion.
8, Emergency Evacuation of the Containment Sphere The licensee is handling' fuel in parallel with the installation ~ work on the reactor depressurization system (RDS). The RDS installa-tion and' fuel' handling activities'can involve thirty or more people ^ in the sphere at any one time. Because of the large number of people that could be in the sphere during the fuel covements, the" inspector discussed provisions for ti=ely evacuation.of the sphere in the event of' an emergency, giving consideration to the capacity of'the air lock and the length of time to pass through the air lock. The licensee completed a preliminary review of this item while the inspectors were at the site. The licensee's preliminary review indicated no unacceptable personnel hazards would result from a post'ulated fuel handling accident.
9.
Emergency Cooling for Fuel Transfer Cask The inspector noted that the step ladder and hose connection used to rig the transfer cask with an emergency supply of cooling water were not designated as safety equipment. As such they are uncontrolled and cannot be assumed to be available when needed.
The licensee stated that these items need not be controlled because the FHSR calculated 200 minutes time period before the water in fuel cask boiled away.-The inspector has no further questions regarding this ite=. , e
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_ . REPORT DETAILS -. Psrt II.
.. Prepared By: -
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C. - Erb-(Date) . . Reviewed By: / N <[ J. C. LeDoux (Date) 1.- Persons Contacted The following individuals were contacted during the inspection.
. . Consumers' Power Company (CP) C. C. Tyson, Maintenance Superintendent S. E. Martin, EngineOr - Maintenance R. Stafford, Inspector - Radiography . H. Keiser, Engineer - Operations-Catalytic Construction Connany (Catalytic) J.-Chapnan, Supervisor - Quality Assurance _ G. Kenny, Quality Control Inspector 2.
Core Spray Valve-Relocation a.
Reason for Change Two isolation valves, a check valve, and associated piping were raised several feet in elevation, so that malfunction of the isolation valves, due to flooding, could not. occur.
This involved renoving about 17 feet of pipe and then re-welding the system.
b.
Materials and Specifications The Powell isolation valves are carbon steel, No, MO-7051 and No. MO-7061, and the connecting piping is four-inch - diameter to Specification ASTM A-106, Grade B.
Wall thickness of part of the pipe is a nominal.237".
Proce-dure No. MPIS-3, Revision 6, cevers relocation of two . - 10 - i , .- .
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' valvas cnd en2 flow element. _ The piping was in talled to the.
requirements of ANSI B31.1 - 1973, Power Piping Cod,c.
The radi- -trnphic and penetrant inspection acceptance were to be based . s Section III, NB5300, while repair was based on Section XI, ' ' li?B4423.
In general, the rerouting was performed to ASME Sec? ion XI, 1971 edition, paragraph IS-400.,- . c.
Qualit? Assurance Results .. Recocds for four velds, No. IA, No. 14A, No. N, and No.'S, were examined. These welds were made using a V-groove preparation.
The gas tungsten are process was used with an open butt to fuse in the root. The welds were completed using the shicided metal arc process. Procedure and personnel qualifications to - Section IX were in the file. - The radiographs indicated many repairs had been made with porosity, due to loss of protective gas and burnthrough which resulted in thin and thick areas of weld. Welds No. South and No. 14A showed the poorest quality. An identification of two welds as North and South in the system could lead to problems, and it is recommended that welds should have a number or letter in the weld number which identifies the system. A number-band, which-located areas around the circumference of the weld, was used, but the paper work for repair did not indicate required grinding areas for veld repair. A baseline UT inspection was_ performed on the above welds, but no indication of root abnormalities was shown.
The prerequisites for the activity were signed off by the code inspection but the final repair including review of - radiographs and ultrasonic testing was not signed off.
3.
Reactor Depressurization System (RDS) An automatic depressurization system for the reactor system is a.
being installed at the present time. Catalytic Construction Company (Catalytic) has the contract to install the headers, valves, and piping for this system. Grinnell Company (Grinnell) has furnished shop-welded spools for this job.
Four isolation valves were supplied by Anchor-Darling Company, and four safety relief valves were supplied by Target Rock, b.
Quality Assurance The inspector examined the radiographs and other NDE records for the following welds and found them acceptable to QAP-7125.
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. Weld Origin F'bricater Siza Pr:cesc Walder- - A-RDS-101-J1N1-5 Shop Grinnell 12" Automatic -
B Shop Grinnell 12" Automatic.
- 102-D-1 Field Catalytic 6" Manual PF-9-101-6 Field-Catalytic 12',', Nbnuoi PF-4 Catalytic is using.the gas tungsten arc process with a Crinnell insert for the wcld root', followed by shielded metal arc process to completion.
Section III, 1974, edition of the ASME . Code, is the governing document. The Catalytic procedures and personnel were qualified to ASME Section IX.
The Class 1 valves were produced with an "N" Stamp affixed, and certifications as to materials were in the Catalytic QA files. Certifications, as to minimum wall thickness, were also in the file and satisfactory.
The inspector understood that CP quality assurance representa-tives are auditing the QA' results of the Catalytic operation.
However, no signatures of CP QA representatives for acceptance were seen on the quality documentation.
4.
In-service Inspection a.
Status Inspection CP has contracted the in-service inspection to Southwest , Research Institute (SWRI) for the reactor vessel and recirculation piping. CP expects to perform examination of the steam drum and supports. U-Tech Company arc'also expected to do some of the inspection work.
SWRI will make as-built isometrics of the piping systems with particular emphasis on locating and inspecting all accessible bimetallic welds.
SWRI or CP will determine the-length of longitudinal welds in vessels requiring a percentage inspection over a ten-year period.
Eight Class 1 valves are scheduled for valve wall thickness determination during this outage.
The inspector was shown an overall inspection plan indicating the number of welds which will be updated when SWRI completes its work. No procedurcs from SWRI had been approved, and a meeting was held on February 5, 1976, at CP corporate head-quarters with the authorized inspector, at which time all procedures and NDE plans were to be approved.
SWRI was scheduled to begin work en February 10, 1976.
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. REPORT DETAILS . Part III . Prepared By: D. R. Hunter -
. 1.
Persons Contacted
. C. J. Hartman, Plant Superintendent . D. E.-DeMoor, Technical Superintenden: C. R. Abel, Operations Superintendent C. C. Tyson, Maintenance Superintendent G. B. Sczczotka, Quality Assurance Superintendcat C. E. Axtell, Chemistry and Radiation Protection Supervisor A. C. Sevener,; Shift Supervisor R. E. Schrader, Instrument and Control Supervisor.
S. G. Martin, Plant Engineer F. J. Valade, Shif t Supervisor T. M. Brun, Assistant Chemistry and Radiation Protection Supervisor . S. A. Carlisle, Shift Supervisor R. W. Doan, Training Coordinator, Shif t Supervisor C. F. Sonnenberg, Assistant Shift Supervisor J. J. Zabritski, Quality Assurance Engineer 2.
Review of Reportable Occurrence Reports P-01-76, inadequate design pressure ratings on a vacuum a.
transmitter and six presrure switches, reported on-January 19, 1976. The licensee reported 17 that during an evaluation of the containment penetrations regarding Appendix J to 10 CFR 50, the vacuum transmitter PT-173; con-tainment pressure switches, PS-664-667; and containment pressure / vacuum switches, DPS-9051 and DPS-9052, would not withstanc'the containment pressure of 23 psig anticipated d 2 ring the DBA.
The inspector reviewed operating memo 2-76, lugged in the-Shift _ Supervisor's log on January 21, 1976, which was in effect until the end of cycle 13.
The containment vacuum transmitter and the containment pressure / vacuum switches were noted as isolated and were required by the operating memo to be unisolated at a containment pressure below five psig following a containment pressurization event. The inspector verified-that the licensee plans to replace the switches during the present outage.
5/ CP to IE:lII, ltr dtd 2/2/76., - 13 - ( r , .. - . -
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Review of Plant Operations , The inspector reviewed the following selected records'of reutine plant operations to verify these activities to'be.in accordance vitt the Technical Specifications ~and Administrative procedures.
Shif t Supervisor' Log, -Jan"uary 8, '1976 through February 2, '1976.- a.
b.
Control Room Operator Log, January 13, 1976 through February'1, ~ 1976.
c.
Reactor Operator Log, January 6, 1976 through February 4, 1976.
d.
Control-Room Data Sheets, January 1, 1976 through February 3, 1976.
(1) The inspector noted that the stack gas monitor had been logged as frozen end no readings taken on several occassions during January.
(a) 1-18-76, 1400-2400, 10 hours (b) 1-19-76, 00-0800, 8 hours
(c) 1-22-76, 1300-1600,-3 hours (d) 1-23-76, 1300-1500, 2 hours (e) 1-24-76, approximately 6 hours The stack gas monitor, including the iodine and-particulate filters were out of service in January-for approximately 29 hours, and a total of 18 hours continuously ~ on January 18-19, 1976. The inspector verified that a QA-16 had been issued (November 25, 1975), and that the I probica was being pursued to correct the apparent freezing of moisture in the embedded lines in the stack between the isokinetic probe and the equipment at the base of the stack.
The inspector verified through discussions with the licensee representative that the filter flows and any activity changes i during January were considered during the outage times, e.
Operating Memos.
f.
Daily Orders, January 6, 1976 through February 2, 1976.
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Control Room Strtus Bor.rd.
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' Outstanding tags t.nd tagging orders.
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Fuel Status Boards.
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Administrative Key Control.
4.
Review of Refueling Operations ~ a.
The inspector reviewed the master checklist and selected . systems checklists-to determine that systems disturbed during the refueling outage will be returned to normal prior to unit refueling and plant startup operations. No discrepancies-were noted.
b.
The inspector's review of activities associated with the fuel transfer cask revealed that the safety cable braking mechanism was not addressed relative to inspection or testing requirements; nor was the inspection or testing of the mechanism included'as . ascheduledprgyentativemaintenanceitem. The licensee's correspondence-with licensing. indicated that the braking-mechanism was the insis for moving the fuel 1 transfer cask directly-from the core to the spent fuel-pit at an elevation of approxi-mately 1 feet above the refueling deck. This elevation above.the floor allowed the 10 inch distance required to activate the safety braking mechanism in the case of a fuel transfer' cask drop accident. The inspector verified that the safety braking mechanis=.had been accidentally tripped during the cask rigging operations immediately prior to fuel covements. The licensee indicated that the mechanism had bcen tripped on several other occasions during the past.. This item will remain open pending-the completion of a review by the licensee to determine the inspection and testing requirements of the safety braking mechanism.
5.
Review of outstandine Items The inspector reviewed selected outstanding items to determine licensee followup actions.
a.
The station battery seismic requirements were addressed in Consumer's Power internal correspondence dated December 19, 1975, in answer to AIR BRP 67-75, which requested an engineering evaluation. The evaluation indicated that the battery was installed in accordance with the Final 6/ CP to NRR, ltr dtd 1/22/76.
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. . ' .Harards Summary Report, section 2.6, and additional support for the battery was not required.. The FHSR.indi- ~ cates the containment,-concrete structure, and the.cquipecnt are designed to at 1 cast maximum ground acceleration rate of - 0.05 gravity. The containment vessel is designed to withstand a vind force on 'the vessel of.'100 miles per hour, which exceeds the earthquake forces.- The equipnent installed outside the containnent vessel does'not appear to'be addressed-in the FHSR..The station battery ~is the power supply for the- -DC-ECCS valves (core spray and building spray valves) and. the power supply for various other engineered and operational safety features, including: (1)' Liquid poison systes controls.
. -(2). Reactor building ventilation and vacuun. breaker valves.
- (3) 480V notor control center 23 control pover.
(4) Protection Bus #3.
(5) 2400v switchgear control'pover.
~ (6) Emergency Condenser outlet valves.
(7)- Control roc = annunciators.
This iten vill re=ain open pending further reviev by IE:111.
b.
The inspector reviewed the off-gas isolation test perforced on January 31,1976, (0-WCS-1). The off-gas holdup line isolated and the pressure increased as expected over the duration of the test. The stack gas activity initially decreased, but ienediately increased again; indicating an unidentified off-gas isolatien syste= bypass flow path.
The condenser vacuu= did not change substantially throughout the test period, indicating that a bypasc flovpath exists.
The inspector reviewed the off-gas isolation with the licensee, which has been an outstanding item since June 1972.1/ This iten vill be followed during a subsequent inspection -trip.
The inspector reviewed the co==unications between the licensee c.
and LTJt concerning the containment vacuum relief -syste=. The inspector noted that the controls for nanusi operation of the vacuum breaker valves and an alternate f1'/w path vere being reviewed by the licensee. This ite= will re=ain open pending completion of the review by the licensee.
7,/ CP to DL, Itr dtd 6/26/72.
(~ 16 - - . . O . .
. _ _ _ - _. - . . . Theinspectorreviewedthefailureanalysispackageassopiated ~ ' - d.
with the' cone.rol rod drive relief valve nipple l failure.d The failure was due to fatigue caused by vibrations associated with the positive displacement charging pumps. The licensee has initiatedLan engineering review to provide' supports for the CRD piping subject to failute by vibrations during CRD
pump operation. No discrepancies were noted., The inspector reviewed the revised. administrative' procedures e.
concerning temporary and permanent procedure changes to operating procepures (1.4.A.6) and to. operations checklists (1.4.A.3.5.1).9 No discrepancies were noted.
f.
. The inspector reviewed with-the operations superintendent' the corrective actions concerning review of the maintenance procedures associated with the control rod drives.as a result of A0 050-155/25-75. The licensee representative indicated that during 1976, procedures for-major maintenance items are planned to.be written.
The inspector reviewej the c {{petive actions associated with g.
the item of noncompliance.]O This item remains open.
-_.
(1) The required ! circuit analysis was not. completed.
(2). The plant review coc=ittee is continuing to deter-mine cach facility change as safety related or not-immediately prior to commencing the facility work until iten (4) below is completed.
, The inspector reviewed selected construction activities management controls to insure the activities (work packages)_ safety reviews were being performed by the ~ plant review committee prior to commencing the specific . work activities.
(a) Sciccted work packages indicated completion of safety evaluations of the plant interfaces.
(b) Field change notices were reviewed by the responsibic engineer and if necessary, due to being outside the original safety evaluation, were approved by the plant review committee.
8/ A0 050-155/29-75.
- 9/ IE:III Inspection Report No. 050-155/75-10.
/ IE:III Inspection Report No. 050-155/75-15.
Jy/ CP to IE:III, Itr dtd jl 12/19/75.
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. . . - . . (c)' The ' selected construction activities were being performed within the approved work packages.
. (3) The Q-list is in effect.
'4) The corporate and plant procedures relating $to modifi-( cation control have not been revised., . 6.
Facility . The inspector toured the facility to view the plant and construction activities, a.
The construction areas were becoming cluttered with work items such as hoses, extension cords, and miscellaneous-items. The inspector noted that the licensee had extra personnel assigned to remove trash from the construction areas.
b.
The inspector noted the use of cicar polyethylene sheeting for general purposes around the spent fuel pit and the open reactor vessel. The inspector questioned this practice based on the difficulty of seeing and retrieving clear poly or clear plastic from the pit or the reactor vessel and the possibility of returning the plant to operation with_ plastic items in the primary coolant system.
c.
The inspector reviewed the operation of the containment per-sonnel hatch electro-hydraulic operating mechanism following the improper operation of'the interlock system on the inner door during containment exiting. The malfunction, 0-ring compression, was apparently caused by the increased usage of the doors during the major construction activities. The inspector verified that the surveillance of the access hatch hydraulic pressure will be increased pending a further review by the licensee and a possible permanent change to the door operating mechanism.
! d.
The inspector reviewed the plant illuminated annunciators with a control room operator. No discrepancies were noted.
7.
Quality Assurance The inspector reviewed with the licensee representative the present status of the implementation of the quality assurance program, i l I - 18 - lI l l l
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The Consumers Power quality assurs, ace policies are:in final review in Licensing at the present time.
b.
The~ licensee plans to establish in.lementation dates associated.
.with'the specific areas of the quality assurance program manual for nuclear power plants.
c.
A selective review of the quality assurance-program procedures-indicated that the. licensee is presently involved in impicmenting program requirements.
d.
The plant quality assurance group appears-to be performing , plant' audits and surveillance on a-limited basis. An audit plan has.been issued but is not yet approved.
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