IR 05000155/1976013
| ML20002D541 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/08/1976 |
| From: | Hunter D, Jordon E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20002D540 | List: |
| References | |
| 50-155-76-13, NUDOCS 8101210321 | |
| Download: ML20002D541 (13) | |
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NUCLEAR' REGULATORY COM:tISS10N OFFICE OF INSPECTION ~AND ENFORCEfENT.
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REGIO III:
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Report of Operations--Inspection
IE Inspection' Report No..050-155/76-13-s l
Licensee:
. Consumers Power Company 212-West' Michigan-. Avenue Jackson, Michigan 49201 Big Rock Point Nuclear Plant'
License No. DPR--6 Charlevoix, Michigan-Category:
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Type of Licensee:
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Type of Inspection:
Routine, Announced Dates of Inspection:
June 18-19 and 21-24, 1976
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i Principal Inspector:
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7 (Dafe)f.
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Accompanying Inspector: None i.
Other Accompanying Perso nel: None
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Reviewed By:
r.. L.
<>rdan, Chief Reac r Projects Section 2
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SUMMARY Op pIND1NGS '
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. _ Inspection Summary
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-Inspect on on~
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' June-18-19'and,21-24,' 1976,-(76-13): Review offpre-
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. operational' testing, startup testing,.' operations,' reportable oc'currence,.
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headquarters-requested item and selected outstanding" items'. Jone item of:
s noncompliance - was identified !concerning. f ailure to - performitemporaryi c
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s procedure: changes 11n accordance with the Technical Specificatibns.
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EnforcementJItems-
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15ua following item was -identified during the inspection:
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A.
Infraction-t Contrary to Technica'1 Specification 6.8.3',' temporary procedure changes regarding containment entry during the Hot;Valvc;0perability
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Test were madeJon June _17'and 18,'1976 without the-required approvals'.
(Paragraph?3, Report' Details)-
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Licensee Action on Previously Identified.Inforcement Items-
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A review of plant modification controls indicates that the licensee's:
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corrective actions are not complete.
(Parat.raph 110,: Report Details)
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Other Significant' Findings:
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5 A.
Systems and Components-
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The startup testing of the reactor depressurization system has l-
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the depressurization valves.
(Paragraph 3, Report Details)
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2.
Unresolved Item: The lack of documentation at the site of'a
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design review concerning the modification to the emergency diesel generator control circuit is considered-an unresolved
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l item;pending further review.
(Paragraph 8.g., Report-Details)
B.
Eacility Items (Plans and Procedures)-
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I The low power physics testing has been completed-following the-
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initial criticality on June 16, 1976.
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Managerial Items None.
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-NoncomplianceIdentifiedandCorrectedib[jLicensee,
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None.
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Deviations
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Status of Previously Reported Unresolved.-Items
'1.
The design review concerning the.electricalf power; supplies.
-forfthe: reactor depressurization' system uninterruptable power.-
supplies has been: completed. : This item is considered resolved.
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-The authorized ' inspector-sign'-off of the werJsiin the ring core-
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. spray line has:been completed..This-iter is; considered-resolved.
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(Paragraph; 5.b., _ Report Details)
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3.
The: authorized inspector sign off of ther1974' inservice-
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resolved. ' (Paragraph 5.c., - Report Details) -
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Management Interview The management interview was conducted cnt 'Juue. 24,1976,:by Mr. Hunter 1
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with the following persons present :
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R. B. DcUitt, Manager, Production C.'J. Hartman,' Plant Superintendent
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D. E.- DeMoor, Technical' Engineer C. R. Abel, Operations Superintendent T. W. Elward, Technical Superintendent -
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- J. P. Flynn, Maintenance Superint'endent
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G. B. Szczotka, Quality. Assurance Superintendent
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A. C. Sevener, Operations Supervisor
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R. W. Voll, Reactor Engineer-V '. Avery, Shift Supervisot A
A.
TF-inspector stated that a review of operations revealed only-one j
questionable item-concerning the' procedural controls for rod withdrawal and insertion in single notch step sequences. The a
i operating ' procedure and the Technical Data Book. did not appear.
to provide clear instructions for the operators.
The licensee stated that the area would be reviewed and the appropriate corrective actions taken.
(Paragraph 2.e., Report
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B.
The inspector stated that containment entries made during the performance of STP-10 on June 17 and 18, 1976,. which were contrary
.to the hot valve operability procedure represented noncompliance
with Technical Specification 6.8.3. which requires approvals of-(
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i temporary procedure changes prior to feplc=entation..ThcElicensee.
stated that temporary procedure changes had been subsequently issued and a deviation report had been issued,and reviewed concerning the failure to follow procedures.
The licensee stated that the Lfailure to make the procedure change was an oversight.
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The inspector asked the-licensee.o insure that all plant departments were made aware of this problen. The licensee stated that.the appropriate plant staff cembers would be informed. No further response is required concerning this item of nonconpliasce.
-(Paragraph 3, Report Details)
C.
The inspector stated that a review of procedure D2.25, Er.ergency Shutdown, revealed no subsequent opefator. actions to nonitor1 plant conditions.
The licensee acknowledged the statenent and noted that considering all equipment cperating with no assumed f ailures, the reactor system would reach a cooled coddition via the nornal operation of the emergency condenser system.
The inspector stated that this ratter would be reviewed further during a subsequent inspection.
(Paragraph 4, Report Details)
D.
The inspector stated that a review of the open item concerning=the uninterruptable r"er supply battery specific gravities and cell replacenent was.erified to be co pleted. The inspector stated
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that he had note. thct the "C" and "D" batteries continued to exhibit specific ravity readings at the lower end of the acceptance criteria. The li ensee acknowledged the statenent by-the inspector.
(Paragraph 6, Rrr.rt Details)
E.
The inspector stated that a review of the facility change packcge concerning the reactor depressurizatics system electrical power supplies revealed no discrepancies. This item is considered closed.
The licensee acknowicdged the statement by the inspector.
(Paragraph 5.a., Report Details)
F.
The inspector stated that a review of the facility change concerning the modification of the energency diesel generator control circuit revealed that no detailed design review was availabic at the site at the time of the inspection.
The licensee stated that the design review was being located and will be provided. The inspector stated that this type of inade-quacy must be prevented in the future with particular attention baing given to facility changes until the design control procc-dures are fully impicnented.
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During a subsequent telecon with the licensee on June 26, 1976, the inspector stated :that this item will br carried as unresolved
.pending location of the design review onsitc to support the
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documented safety evaluation.
(Paragraph 8.g., -Report Details)
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REPORT DETAILS
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Persons Contacted C. J. Hattman, Plant Superintendent D. E. DeMoor, Technical Engineer C. R. Abel, Operations Superintendent T. W. Elward, Technical Superintendent
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J. P. Flynn, Maintenance Superintendent G. B. Szczotka, Quality Assurance Superintendent R. W. Voll, Reactor Engineer A. C. Sevener, Operations Supervisor R. W. Doan, Shift Supervisor, Training Coordinator E. F. Peltier, Shift Supervisor S. A. Carlisle, Shift Supervisor R. A. Curtis, Control-Room Operator.
H. E. Downing, Control Room Operator D. J. Horstnan, Control Room Operator W. J. Woods, Control Room Operator J. L. Kuemin, Plant Engineer D. D. Herboldsheimer, Maintenance Scheduler H. M. Phelps, Assistant 16C Supervisor G. H. Petitjean, Plant Engineer S. E. Martin, Project Engineer W. Clark, Projects Construction Superintendent
K. F. Krueger, Consumers Startup Engineer A. J. DeGrasse, Catautic Startup Coordinator 2.
Review of Operations The inspector reviewed the following selected records of routine plant operations and conducted plant tours to verify activities to be in accordance with Technical Specifications and Administrative Procedures:
a.
Shift Supervisor Log, May 31, 1976, through June 18, 1976.
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Control Room Log, May 8, 1976, through June 18, 1976.
c.
Reactor Operations Log, April 27, 1976, through June 20, 1976, d.
Control Room Data Sheets, May 5, 1976, through June 23, 1976.
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Operating Memos, May 26, 1976, through June 17, 1976.
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Control Room Status Board.
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Outstanding Tagging and Tagging Orders.
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Fuel Status Boards.
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Equipment Rotation.
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Plant Annunciators.
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The peditiun of selected valve positions =on.the ' firc :. ~
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protection ~ system includingjthe suphly" valves to theLcore 1 ~
sprayfsystem,-emergency ~ makeup to the; main condenser and the hose connection for thE7 core spray'recirculat' ion. heat'
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cxchanger alternate cooling-water supply were' examined;
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Lduring a plant tour.
1. y Technical: Data' Book
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'The review of the icchnidal.Daia Book,.. Technical Specification'5.2.6,
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Tand Operating' Procedures, Sections Bl.l.6.-and B1'3.3.2,concerning.
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control' rod' withdrawal and' insertion sequencir-quirementsi
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i indicated a: procedural weakness. The inspa rified;
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through~ observation ofLrod movement by_' opera
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with' operators that rods are moved.in. single-steps within1the,
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groups following a mirror ~ image pattern'ulth'no rod more than=
I one step' apart ~.7 The procedural steps do not appe'ar: to.be definitive enough to. prevent "an > operator deviation from'the:
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_ rod sequence within'a 'spacific group of rods.
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Operational Surveillance-Tests'
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The inspector reviewed selected ~ surveillance tests performed
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during the outage. _No' discrepancies were noted.
i TR-06.- Liquid Poison System Check Valve Test ~, performed on February 11, 1976.
TR-08_(365-05) - Core Spray System Check Valve' Test.
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TR-09 - Core Spray Heat Exhanger Shell Side Flow,' performed L
on June 1, 1976.
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This test was. performed.to verify' adequate flow through the temporary hose connection between the'
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fire protection system test header and the CS heat exhanger:. The. minimum. required flow was
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131,425' lbs/hr and~the conservative' calculated flow was. 162,000 lbs/hr at a fire pump discharge pressure of 140 psig.
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TR-16 - Emergency Diesel Generator Auto Start, performed-on April 8, 1976.
TV-10, Rev. 2 - llydrostatic Tcst of NSSS, performed on.
June 14, 1976.
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Startup Checklists The inspector reviewed selected startup. checklists and valve checkoff lists, including the Master Checklist, incore instruments, neutron monitoring systems, condensate system, control rod drive (
system, post-incident system, cmergency condenser system, fire protection' system and plant-locked valves.
No discrepancica were noted.
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Reactor Depressurization Startup. Testing of Program 3.
The inspector reviewed the procedure-for startup testing of the reactor depressurization' system (RDS) and made direct observations of the plant conditions during portions of the testing.
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The.O-RDS-l (STP-10), R'ev. 2,110t Valve Operability Test, was-performed partially on June 16-18, 1974 at 100 psig and 700 psig.
The licensee found that the RDS valves leaking through slightly and the isolation valve open limit switches wert damaged due to apparent increased travel at higher system pressure. The test was terminated on June 18, 1976, and the plant was placed in the cold shutdown condition. The procedure was evaluated and revised to allow testing at 1350 psig to provide adcquate pressure to lift the RDS valves via the bypass line.
The inspector noted on June 18, 1976, that maintenance and inspection was being performed inside he containment vessel with prescure in the reactor vessel. These activities were not in accordance with the procedure, step 2.2, which stated that "No entries will be made while there i.: pro-sure in the reactor vessel, except to adjust CA-135.
Entry under this condition will be with the power to the valves tagged off at their respective breakers." Record review revealed that an entry had also been made on June 17, 1976, for.
maintenance purposes. Fo temporary proceduce changes were issued to provide immediate evaluation of the activities.
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Failure to provide the required temporary procedure changes is considered an item of noncompliance pureusnt to Technical Specification 6.8.3.
Following identiff tion by the inspector,the licensee issued temporary procedure changes oncerning the-deviation from the procedure and issued a deviation form (QA-16) concerning the failure to provide the temporary procedure changes as required by.the Administrative Procedures.
The licensee recommenced the startup testing on June 20, 1976, and heated the plant to approximately 1200 psig in accordance with procedure 0-RDS-1 (STP-10), Re.. 4.
During the test at appro::i-mately 500 psig and 1200 psig, problems were encountered with the Target Rock Valves. The "A" and "C" valve opened and closed normally upon demand. The "B" and "D" valves malfunctioned, f ailing to open The upon demand and f ailing to close upon demand, respectively.
plant was placed in the cold shutdown condition to await an evaluation and resolution of the valve malfunctions.
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Energency Shutdoun Procedure D2.25 The inspector reviewed the procedure to provide instructions for shutdown of the plant from outsideftho' control roon. The procedure required the. operator to proceed to the No. 1.and No. 2 reactor protection buses and open the protection breakers (CE 40Al and CB 40A2). The procedure does not indicate any' subsequent
' operator actions to nonitor the plant conditions following the remote trip, but indicates that the emergency. condenser will go into service.
5.
Previously Unresolved Items The inspector reviewed previously unresolved items to verify corrective action completion by.the licensee, The design review of facility change'FC-351 concerning the a.
electrical power load addition to the 1A and 2A electrical buses for the reactor depressurization system was reviewed.1/
The design review was performed by the licensee and the package included.the bus lead requirements and breaker tripping requirements.
No further questions are required at this time and this item is considered resolved.
b.
Core Spray Wei o The revie. __ the core spr'ay weld (North, lA and 14A)
packages /3/, indicated that the authorized inspector reviewed
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and signed off on Iby 6, 1976.
This item is considered resolved, c.
Inservice Inspection The review of the 1974 inservice inspectioni/1/ ndicated that t'h e i
authorized inspector signed, off on June 14, 1976.
This item is considered resolved.
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IE Inspection Ept No. 050-155/76-12.
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IE Inspection Rpt No. 050-155/76-09.
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Ltr, Cp to IE:III, dtd 5/26/76.
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IE Inspection Upt No. 050-155/75-13.
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IE Inspection Rpt No. 050-155/76-10.
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Outstanding Itens'
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The inspector's review of the startup test procedure-package, a.
STP-011,6/ UFS.Functienal Test, revealed that. the specific -
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gravities for all of the battery cells had been taken and recorded, the UPS battery cells D-3 and D-20 had bcen replaced and all of the cell specific gravity readings indicated 1.200 or greater. The inspector noted that the "C" and "D" batteries continued to exhibit specific gravity readings at.the Icver end of the acceptance criteria.
No further questicas remain at this time and this iten is considered cles(d.
b.
The inspector,'s revicv of. the e=ergency procedure, training and walk throughs,7/ indicated that the required training was completed and the shif t supervisors and operators who hal-rissed the training verc being updated pricr to asturing shif t respcnsibilities.
No further questicas are required at this tine and this itcn is.
censidered closed.
The inspector's review of the operatiens training subject naterial I
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covered during the cutage to update the cperatcrs concerning the EDS nodification, ECCS - nodifications and other essential
- facility and procedure changes appeared adequa';e. The shift supervisers and operaters who =issed the trair.ing vere being
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updated prior to assuring shif t responsibilities, The inspector's review of the use cf the functionally d.
Equivalent Substitution (FES) renesEI revealed that the Plant Review Connittee had reviewed the recos. The plant Superintendent has assigncd the task of writing an Ad=inistrative Procedure to control the use of the FES me=os.
This iten vill remain open pending cenpletion and review of the ccepleted Adninistrative Procedure.
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IE Inspection Rpt No. 050-155/76-12.
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1E Inspection Ept No. 050-155/76-07.
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IE Inspection Rpt No. 05C-155/76-10.
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The ins cctor's review of the control of the isolation valve 9/_0_/ for the differential pressure switches (dps/9051 and dps/9052) and the vacuum indicator (PT-173)
indicated that the velve will be normally closed during normal plant _ conditions and vill be unisolated following a'LOCA as the pressure decreases below 5 psig and reisolated any time the containment pressure exceeds 8 psig. The instructions for operation of the root valve located at penetration H-96'are included in step D3.3.2.4 and Appendix B to D3.3 (Loss of Reactor Cog 4 ant).
No further questions are required at this' time and this iten is considered closed, f.
The inspector's review of the surveillance rocedure, for.
the control and indication slow blow fuses,_1/12/ revealed that the surveillance test (T30-23, Rev. O, 6/7/76) is scheduled to be perforced conthly.
The inspector verified that the fuse blocks were labeled and that the operator on duty was familiar with the fuses and the surveillance test.
This iten is considered closed.
7.
Startup Testing Af ter Refueling The inspector reviewed selected.startup procedures to insure selected tests were performed in accordance with the Technical Specifications.
IR-21 - Control Rod Drive Fricticn Testing Procedure, perforned a.
on February 6, 1976.
b.
TR-46 - Core Load Procedure, performed on May 1, 1976.
TR-43 - Shutdown Margin Check, perforced on May 3, 1976.
c.
The inspector verified thot the :est would be terminated if the predicted rod movement would take the reactor critical.13/
next For Cycle 14, the single rod stuck shutdown margin test was performed satisfactorily, d.
TR-44 - Moderator Temperature Coefficient, performed on June 17, 1976.
No discrepancies were noted.
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RO 050-155/01-76, 10/ IE Inspection Rpt No. 050-155/76-04.
11/ IE Inspection Hpt No. 050-155/76-12.
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13/ IE Inspection Rpt No. 050-155/75-11.
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Facility Changes -
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.I The. inspectori reviewed - selected facility, changes to' insure they
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were performed. in accordance _ with ' the Administrativb Procedures,-
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- Technical' Specifications and-10 CFR 50.59.
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CIS-76-FC-349 -~ Addition of the.resinLsluicc111ne manual ~
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grading; the resin,
= isolation' valve and up/M /.
sluice ~.line valves M
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No discrepancies.were noted, b.
CIS-76-FC-328'- Addition--of a tell-tale-drain on~the resin'
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l-sluic e. linc.16 /
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- No discrepancies were noted.
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' SPS-76-FC-3591-Addition of RDS control panel pouer supply 1
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No discrepancies were noted.
d.
SPS-76-FC-338'--Bus 2B extension.
No discrepancies were noted.
c.
- PIS-76-C-376 - Addition of core spray fl'ow recorder on the t '
' control panel.
' No discrepancies were noted.
f.
PSI-76-C-377 - Providing electrical switching for core. spray flow to the recorder on.the control panel.
No discrepancies were noted.
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SPS-76-C-358 - Providing _125V DC for the emergency diesel'
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generator control circuit from the RDS uninterruptable power supply "A'.'.
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The review of ' the facility change revealed that the of fsite
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design review was not contained within the package...A memorandum from 'of f site to the project engineer -in' icated d
l that the design revice had been performed.
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The lack of a documented design review at the site will.bc carried as unresolved pending arrival and review of the design
review documentation onsite to support the documented safety i
evaluation.
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14/ A0 050-155/19-75.
!p 3 / IE Inspection Rpt No - 050-155/75-15.
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Reportah]c Oc:urrence.
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The inspector reviewed the following reportable occurrence to assure adequate review, evaluation'and reporting.
LER RO-9-76 - tailure of the Emergancy Diesel Generator Breaker to Close Upon 1.oss of Power-Th'e licensee reported 2/ that the emergency diese1' generator-l breaker failed to close upon' loss of power to the 2B bus..The-inspector-reviewed the event with the licensee's repreacntative and verified that subsequent inspection and testing of the emergency diesel generator breaker revealed no discrepancies.
The breaker.and circuit has performed satisfactorily subsequently.
No discrepancies were noted.
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17/ Ltr, CP to IE:III, dtd 6/9/76.
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