IR 05000336/1990019
| ML20058A988 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/15/1990 |
| From: | Baunack W, Bessette D, Bettenhausen L, Finkel A, Oliveira W, Stewart J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058A983 | List: |
| References | |
| 50-336-90-19, NUDOCS 9010290339 | |
| Download: ML20058A988 (13) | |
Text
{{#Wiki_filter:c- ' e ,. s
. ,
.. . ... - .<
i- ! . . , ' >I-U.S.NOCLEARREGULATORYCOMMISSION !
REGION I
i L . t [ Report No.: 50-336/90-19 i Docket No.: 50-336 - - License No.: DPR-65 Licensee: Northeast Nuclear Energy Company P. O. Box 270 Hart.*ord, Connecticut 06141-0270 I Facility Name: Millstone Nuclear Power Station, Unit 2 i Inspection At: Waterford, Connecticut Inspec. ion Conducted: September 10-14, 1990
Inspectors: 64t4A1Ao O W. H. Baunack, Senior Reactor Engineer Date - ' . Operational Programs Section < Operations Branch, DRS l0 O A. E. Finkel,.54nior Reactor Engineer Sate Operational Programs Section L Operations Branch, DRS /o/h)Yfu W. Oliveira', Reactor Engineer D&te ' Operational Programs Section Operations Branch, DRS ' , h} f '/ 8 /10- / J'S. Stewart,- Senior Project Engineer / /- -Date t-s Projects Branch No.-4,-DRP [;, b-heamLeader: [aerl OnAlk - 7[O[?v ~ -0. E. Bessette, Chief-Date b . <-- Operational Programs ~Section m Operations Branch,~DRS-kOOOO g6 . ~ - - I :; iG b' 1. > 37 . i' n
_ -_ _ _ __-_ _ __-__ _ - _
j - , , .... /c ... M !
er.
Approved by: _ __L. 4 Bettenha~usen,' thief Date "
Operations Branch. DRS Inspection Summary: Routine Announced Team Inspection on September 10-14. Ws0 p (Inspection Report No30TfB/90-f9).
~ Areas Inspected: The inspection examined the licensee's actions with respect u - to: (1) open items from NRC Inspection Report 50-336/89-13; and (2) a number [, of allegations subsequent to NRC Inspection Report 50-336/89-13.
Results: No deficiencies were' identified.
The licensee's. actions to resolve the issues were acceptable, thorough, timely, and reviewed by QA/QC and Safety / . Administrative audits.
i , f.: .. Y .. p b
r r ' r = I . , .. ' .s . t U- [c h .. , d T-f[ ' , ,, , ,
' t .ji
, ,.. ..'- F.,
I' . !. ' ,,. .. '
L L - F DETAILS , 1.0 Persons Contacted
- T. Arnett Instrument and Control (I&C) Engineer-Unit 2-
,
- J. Becker I&C Manager-Unit 2
! R. Bonner, Engineering Supervisor-Unit 2
G. Closius, Plant Quality Services Supervisor i
- J. Criscione Director Staff Engineer-Unit 2 t
B. Duffy, Engineering Supervisor-Unit 2 ! ~ J. Humphreys, Maintenance Engineer-Unit 2 ,
- J. Keenan, Director-Unit 2
- J. Laware, Assessment Services Technologist
!
- J. Riley, Maintenance Manager-Unit 2
,
- S. Scace, Station Superintendent
! J. Smith, Station Services - P. Smith, 1&C Supervisor-
- W.-Varney, Planning Quality Services Manager K. Wallace, Engineering-Unit 2 l
t-U.S. Nuclear Regulatory Commission f
- D. Haverkamp, Chief DRP Section 4A
} .
- E. Kelley, Chief, DRP Technical Support Staff
!
- W. Raymond, Senior Resident-Inspector o
- P. Habighorst, Resident Inspector
> i
- Denotes those attending the exit meeting.
.The inspectors also~ contacted other administrative' and technical personnel during the inspection.. 2.0. Eva1Uation of Corrective Actions to IR S0-336/89-13 l 2.1 f(Closed)_ Violations 50-336/89-13-01, 50-336/89-13-09 Procedure Adequacy and Adherence , ! Certain procedures could not, in)some sections, be performed as written., l and procedures were-performed without making. appropriate cha'nges.
Further, l a change was made.to.I&C procedure 2419A during an in progress refueling -;
- activity which changedLthe intent of the procedure.
No documented change - -was made to the procedure, nor was prior.PORC. review or Unit 2 Superinten-
dentJapproval obtained..- .To ensureithat Mil.1 stone Station personnel understood the management.
, ' position on safety work and'other nuclear safety matters,.the following
documentation has been' issued by senior management. over the last few- ! , years: (1)-Operating. Philosophy, July 31 1987;-(2) Safety Evaluations, _, July 22, 1988; (3) Retention of-Safety-Related Evaluation Documentation.
-October 16, 1989; (4) Expression.of Nuclear Safety Concerns, ' .. Aprilj l8,;1990; and (5) Focus on Safety and Teamwork, June 6,1990.
$ k .-
__ _ _ _ _ _ __ _ __ _ _ _ _ - _. _ _ -. _ _ _ __- ___ ? ' . t ' ..
. . L
e i J i E , The information contained in these documents has been presented to the
' site personnel. Discussion with first line supervisors, maintenance.
' electrical, and I&C personnel indicated that they were aware of management positions on safety matters and adherence to procedures.
Additionally, supervisors hold weekly meetings with their personnel
where related subjects are discussed.
Discussions with craft personnel , verified that they were aware of management direction on adherence to procedures and they understood the latest guidance on this subject and that they attended meetings where their senior management discussed the , , subjects of safety evaluations, method of documenting test results, and procedural adherence.
>
- -
To ensure an independent audit of the effectiveness of their programs, the licensee issued Millstone Administrative Procedure (MAP) 5.20 defining requirements for the Millstone Station audit program, which is performed , by the division managers. A listing of managers and audits they are required to perform was issued May 18, 1989 (MP-13096).
The results
of these monthly audits are reported quarterly to senior management.
- t The inspector reviewed the departments' Safety Observation Reports (SOR) for June, July, and August 1990 and determined that documented findings of the 50R's were entered into the licensee problem reporting system.
The inspector also reviewed the March 28, 1990 (MP-90-303) " Millstone t . Station Work Practice Compliance Program" report. The. inspector verified i thet senior manacement discussed the results of these" reports during the ' Millstone Superintendents meeting. The results of these meetings include i findings identified and entered into a tracking-system which lists find-l j ings, due dates, and responsibilities for closure.
The Procedure Upgrade Program (PUP) is in varying stages of completion for , the different Millstone 2 departments.
The inspection focused on the 1&C . department program'which includes-an extensive technical review, rewrite ' and format upgrade, independent verification, and walk-through by crLf t , i personnel.. The program is approximately 50% complete for technical ! specification. surveillance procedures. All procedures reviewed by the " < Plant Operation Review Committee are scheduled to be completed by 1992, i The I&C staff has made numerous procedural enhancements and technical > f changes to procedures not fully through the PUP. Two procedures, not yet through PUP, reviewed by the. inspector reflect the need for. procedure revision or upgrade: 1.
SP-2404A0 " Spent Fuel Pool (SFP) Area Radiation Monitor (RM) Cali-bration" Revision 3, Change 4.
The procedure, which was last revised in January 1989, concerns annual calibration of the spent fuel pool-radiation monitor. Change 1 March 1989, corrected an omission in the revision.
Change 2, June 1989, adapted the' procedure to the writer's guide format (Procedure. Upgrade), but did not include full-technical review and walk-through.
Change 3, June 1989, clarified .the test setup. Change 4. October 1989, rearranged a procedural.
step to allow enhanced procedure compliance.
1 o n
n + c y ,, : [' n) ".1
- j.i
'1 ' w
-
- Di '
h-
- 2.,
SP 24011 " Local Power Density. Test" Revision 7, Change 3, is a , monthly, at. power surveillance.
Revision 7 was completed in e April 1988 prior to-full implementation of PUP. Change 1, June 1988, ' f clarified procedural steps.
Change 2, July 1990, ef fected steps
.H directing the technician to record digital voltmeter calibration and L reference voltages on the data sheet.
Change 3, July 1990, involved L test equipment selection and removal.
L - The length'of time since.the last revision and the large-number of changes . C reflect the need for these procedures be revised.
The licensee intends to ~ ~ a review these procedures and others that have had format upgrade, but no g - thorough upgrade including walk-through, by the end of 1991.
H The following procedures were reviewed that have been through PUP, techni-cal specification review, -independent verification, and walk-through, [L . .
c-1.
SP24010." Reactor Protection System Matrix Logic and Trip Path Relay-i Test" Revision 9; The revision was issued in May 1990 for this l startup test.' i o + ' -
2.
SP2401F." Reactor. Protection System High Power Trip Test" Revision 7 i . Change 1.
The; test is conducted monthly at power.
The revisi.on was ' issued =in September.1989.. The. change clarifies a small number of
' active steps and relocates a " note" within the procedure.
The change ,=j L was processed correctly as a non-intent change on December 7. 1989 ! ' and reviewed by PORC'within the required 14= days on December 13. 1989.
j s . The licensee-completed attechnical! specification compliance verification .! of allEl&C surveillance procedures in January 1990 as corrective action:
Eto-LER 89-008. LProcedures were sampled to 9ntsre that both I&C and
maintenance procedures had; biennial reviews >.s required for 10 CFR.50 l ' JAppendix B.c -All procedures checked had=been reviewed as2 intended.
1 . L> =The team. discussed procedural adherence with a number of 1&C technicians.
- l
& 'A good attitude of procedural compliance and making. changes when required; a > 1
was evident. An. observation 'of SP-2401Q " Response Time Test-of_ Reactor 'l Le i ~ Coolant RTDs" was conducted and procedura1' compliance was evident. The 'i t . technician performing;the surveillance _had processed a' change earlier in.
F* the day,: before the surveillance was conducted..tof reflect a; change in: thel . , W f
- test. equipment being used,and war observed following: procedure steps for'
. ttime response runs. , .
p Finally,cLicensee : Event' Reports.'(LERs) were1 reviewed to. determine if any - }i ' ~ ' k reportable occurences couldlbeLattributed tofinadequate procedures or:~ i B - lack {ofiprocedural compliance; The review consisted of aLeheck;of 1989E .i b Land 1990-LERs through.90-12.; LER 89-09, wh_ich reported an 1noperable.
~ W .
- radiation monitor RM-8262,3was. caused when a change to a surveillance
.
P, Sprocedure was prepared improperlyLand inadequately reviewed. Otherwise;
(no evidence existed to suggest that' station events were.being caused by: j; , -
- procedural inadequacies or procedural non-compliance.
i t .
. .
.
' f
. j , , , $[
fe ?-
,
_ . c , ,
.. .-
L The licensee's actions to notify their personnel of management's position on safety matters, revising and upgrading procedures, and following proce-dures as written is adequate and is associated to the level of safety associated with these issues.
These items are closed.
2.2 (Closed) Unresolved Item 50-336/89-13-02, Non-conforma $eReporting _ The licensee had not resolved the internal non-conformance reporting audit findings.and the control of inputs, such as the informal three part memo, to the formal reporting system.
~ Procedure NEO 3.05, Non-conformance Reports, was revised and issued in =D January,1990, followed by an in process verification (IPV). An IPV-3005 l report of February 9, 1990, closed the concern regarding non-conformance reports (NCRs). A QA audit started on September 14, 1990, to review the effectiveness of the NCR effort.
Corrective Action Request (CAR) 90-01 issued.on January 29, 1990 included the Plant Incident Report (PIR) concerns., The CAR was closed on April 12, 1990, and the PIR effort will be audited in mid 1991.
ACP-QA-1.20, Problem Reporting, was issued on December 1, 1989, to provide guidance on the use of various problem reporting systems avail-able to. individuals working at Millstone, it also formalized the three c part memo reporting system and requires a 10-day initial feedback for logged entries. The inspector reviewed AO-QA-1.20 and discussed the use of the three part memo with licensee personnel. A monthly meeting-was held on January 5, 1990, where the I&C Department Head described - l the application of ACP-QA-1.20. The I&C department has a system for receiving. investigating, reporting, and tracking the three part-
.
' memo system in a satisfactory and timely manner.
Based on the closing of the audit findings and the satisfactory implementation of the three part memo reporting system, this item is closed.
. 2.31,(Closed)Unresolveditems 50-336/89-13-03 and 90-02-01 AVMS Procedure ~ Acoustic Valve Monitor System (AVMS) Functional Test procedure . SP 2410A steps 6.1.6,3 and 6.1.6.4 were difficult to accomplish exactly- , [j as indicated.
Before the July 1988 revision of the procedure,18KHz was used as a reference resonant peak in the noise spectra.
y ' The AVMS is used to detect opening or leaking of power operated relief valves (PORVs).. Each AVMS: consists:of an accelerometer mounted on the discharge pipe. downstream of the PORV, a charge amplifier mounted onLthe ' pressurizer. blockhouse inside the containment, and a signal conditioner < i: in.the control room area. :The accelerometer senses vibration of-the pipe.
The vibration signal is amplified by the charge amplifier and transmitted-toLthe-signal conditioner.
The output from the signal conditioner can be , ! . l ' U Ci _ > ,
_ __.
. _ _ _ _ _ -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ y..' Iy
. , ' '
m I connected to an oscilloscope or a spectrum analyzer. The AVMS's spectrum analyzer measures signal amplitude versus frequency (two dimensional signal) and plots on a printer.
As required by the monthly functional test, the I&C tect.nician was required to compare the newly plotted spectrum with a reference spectrum to determine the acceptability of the test. Station Pro;edure SP-2401A, Revision 4, Change No. 4, November 28, 1989, was issued to delete the requirement for the technician to interpret the spectra.
The technician is to plot the spectrum and out-of-specification results are documented, " while the plotted spectra were evaluated by the l&C supervisor. Change No. 5 to procedure SP-2410A will transfer this analysis to engineering.
The inspector reviewed the proposed change to the procedure prior to PORC/SORC board submittal.
Discussions with both 1&C technicians and supervisors indicated that the proposed change No. 5 to the procedure resolved the technicians' concerns.
During original installation of the AVMS, the procedure required that an 18KHz tone should be present in the collected spectral data.
The 18KHz tone has intermittently failed to appear as early as 1984, according to the AVMS maintenance history.
The original accelerometer had been changed to a crystal resonance different from the installed equipment. Also, the original. procedure SP-2410A was prepared when acoustic testing was being developed for the use in system evaluations and had some basic assumption errors in the use of signal data.
In 1988, the SP-2410A procedure was changed to use a 75Hz tone generated by the reactor pumps (shaft rate at 15Hz x 5 impeller blades).
Since the four' reactor coolant pumps run continuously at full flow under normal plant operating conditions, the tone should be at a constant frequency and at a more or less constart amplitude (Ref. Licensee Letter No.1303909, ' July 1,1988)..The AVMS system signal. includes such inputs as instrument hammer, frequency. response of the piping system and the AVMS transducers.
Because of this complex signal the'licensea directed the evaluation of the AVMS signal to the reliability engineering-organization.
- The AVMS is a method used to determine the operation of a valve using the
. acoustics.of the system.
The data collected by the SP-2410A procedure is evaluateJ by the reliability engineering organization and used as another predictive tool in evaluating the function of selected system valves.
The . AVMS is used as a predictive method of evaluating the performance of valve and the failure of this test system would not effect the safety function-of.the system.
'The procedure change to use the frequency of the reactor coolant pumps 'and the analysis of the' test results by engineering personnel ' resolve the ! original concern with this document. These items are closed, r . ' l >
- .. , . .. ...
1; i 2.4 (Closed) Violation 50-336/89-13-04, Non-Seismic Conduit No seismic qualifications were conducted for the new conduit installation F to the Reactor Coolant Pump (RCP) speed sensors.
i ... Plant design change record (evaluation) PDCR(E) MP2-80-:14-1226P was performed to seismically qualify the new conduit instaliation for the RCPs speed sensors.
Revision 1 of the PDCR(E) included a seismic analysis that confirmed the original installation was not originally seismically , ,, qualified.
The inspector audited the calculations and also verified that ' the analysis was in accordance with the criteria established by the FSAR Section 5: Nuclear Engineering and Operations (NEO) Procedure 5.06, Design Analysis; and NEO Procedure 5.19, Seismic Qualification Reviews.
! This item is closed.
2.5 (Closed) Violation 50-336/89-13-06, Outdated Drawings Licensee fai_ led to control drawings to assure that the most up to date version of drawings were available at the work stations.
L The inspector reviewed:.I&C Instruction 1.10, I&C Department Instrument Loop Folder; Instruction 1.09,.Use of Drawings; and Instruction 1.08, Vendor Technical Manuals, to ensure that procedures have been-documented.
-The licensee replaced the I&C stick files with P& ids in clearly identi-fied file drawers, and trained their technicians to use the Generation Records Information Tracking System (GRITS) to determine the latest i applicable drawing (s).
Safety / Administration audits by I&C supervision and management are conducted.in accordance with Millstone Administrative '! Procedure-(MAP) 5.20.
Attachment 8.1 includes attributes to review
.
drawing revisions.
The inspector randomly selected a number of drawings ! from the loop folders and confirmed using GRITS that the latest versions t were on file.- , Based on the instructions reviewed, verification of the I&C and Main- ' tenance drawings against GRITS and Nuclear Records Offices files, and ' discussions with the-craf t and supervisory personnel, the inspector , concluded that the licensee is controlling their drawings satisfactorily, e This'1 tem is closed.
'
- 2.6 (CloseJ) Violation 50-336/89-13-08 Control of Overtime-y
, c . Licensee failed to control overtime in strict accordance with Generic h ' Letter 82-12.
. Improper control was caused by: (1)'the failure of supervisors to
- consider overlapping 48 hour periods of work wherein the individual.
' exceeds the 24 hour limitation for that period; and (2) failure of ' the unit manager to approve cases exceeding the limitations of Nuc. lear Engineering and Operations (NEO): Procedure 1.90, Overtime Controls for ' ' Personnel Working at the'0perating Nuclear Station. Additionally NRC
, h. ' u
, - - - . - - - -. - - _ _ - - - - - - _ _ _ _
I ! i .,- .- ,.- , i'
i ! ' E letter of April 19, 1990, disagreed with the licensee's initial response I of December 8, 1989, in interpretation of the GL 82-12 guidelines regarding i controls for any 48 hour period.
l ! The inspector reviewed the draft revision to NEO procedure 1.09 and L found that it included the changes to conform to Generic Letter 82-12.
The inspector determined by reviewing overtime records that the adaed i ,, }, controls are being implemented.
Discussions of overtime controls with ! craft, supervisory, and management personnel showed that the personnel r are knowledgeable of the limitations of overtime.
l This item is closed.
2.7 (Closed) Unresolved Item 50-336/89-13-10, Pressure Gauge Hysterisis l An engineering basis was required to support c wrent calibration practices l defined in I&C procedure 1104A, "I&C Pressure Test Gauges Calibrations."
l I&C procedure 1104A, Revision 7 was rewritten and issued on June 19, 1990.
The procedure provides control of permissible error during calibration of' l n pressure / vacuum gauges by placing a limit on the absolute value of the i difference _between the increasing as-found reading and the decreasing
as-found reading.
Steps of the procedure are writteh to ensure that } equipment calibrated using this procedure will not have unacceptable ' errors in the increasing / decreasing reading areas.
, . The inspector reviewed Revision 7 changes with both the I&C supervisor , and technicians.
They have been trained on the revised procedure and ! understood the changes in both the method of testing and verifying the i increasing / decreasing readings as described in paragraph 6.5, " Check for Hysteresis."
l The change to'the~ procedure and the training engineering provided to the
I&C personnel have resolved this testing question. This item is closed.
2.8 --(Closed)-Unresolved Item 89-13-11, Gauge Calibration - ' The same dead weight tester was being used-for both fresh _and salt ~ - water gauges, thus impurities were introduced.
To ensure that salt water and. fresh water gauges are tested with -
. ' independent test equipment to maintain' purity, the licensee' implemented .the fol_ lowing actions: J The MET-LAB'has dedicated 1 oil, 2 fresh water, and 1 salt water
dead weight testers', which are labelled. clearly.
- Th1 distilled water used.in' the 2 fresh water dead weight testers
is changed every six months'. , J - }
-- -
_ _ _ _ _ _ _ - - - _ _ _. .. c - ". l (,.
t .
[ p m ! i j
Maintenance of the dead weight test equipment has been entered
into the Production Maintenance Management System (PMMS).
The. inspector interviewed both the MET-LAB and PMMS pe m nnel and determined that they were knowledgeable of the dead weicht tester
, concerns and that the calibration schedule for this ea m ment has ! 'been entered into the PMMS.
Examination of the PMMS verified that the dead weight testers were on a six month maintenance schedule.
! ! To determine if the system had become contaminated by.the gauge test f ~ equipment, the chemistry department performed an analysis on the volume ,. of contaminated fluid available in the pressure gauge when injected into , p the most restrictive system (feedwater).
The analysis indicated a mini. . l seule value (approximately 0.0036 ppb) below limits of detectability and l L well below the limit'of control for the Steam Generators which is 20 ppb.
. > Further investigation by NUSCO Nuclear Materials and Chemistry determined ' that the instrumentation sensing lines will sustain no detrimental effects ! y~ from the maximum potential quantity of potentially contaminated gauge ' water.
The two documents prepared by the itcensee to support their-analysis are memo MP-SC-89 234, 18 July 1989, I&C test and MP2-1-1381,
' 20 July 1989.
Review of the referenced calculations agrees with the l licensee's response,
a
The control of the MET-LAB test equipment and the additional salt water- ! gauge testing equipment should ensure that the above problem has been . corrected. This item is closed.
-; x
[ 2.9-{ Closed)UnresolvedItem 50-336/89-13-12, Use of Blue Tags.
t ' > L .Two blue tags-were placed on the same system, raising a question about [' tagging procedures.
( > t ' The licensee evaluated the situation described in IR 50-336/89-13and concluded that two blue tags used in the same circuit.at the same time
_is neither an operational nor a safety problem.. Tag configuration issues
- are1 resolved by either the field personnel hanging tags or the job super-
,
visor. verifying tagging. While electrical shock could hypothetically occur- + p if electrical potential was being applied by test equipment-to circuitry [ isolated by a blue-tag, the hazerd and the likelihood of occurrence would
g 'be,the same if the-upstream isolation was a red danger tag a.
.; e , y
- This item is closed.
' i:. I , k',
li ' ? - . ,
, ._.
_ __ .' , ' g s; y '
- .
, o )) , 3.0 Issues Sub'equent to IR 89-13 A number of allegations made to the NRC originating subsequent to the
inspection described in IR 89-13 were reviewed by the 95pectors: ' 3.1 Area Radiation Monitor Information in the FSAR Needs Cch ecting . During routine licensee activities certain errors were identified in the + FSAR concerning local indications and alarm functions of the containment i high range area radiation monitors, and two area monitors which should ~
, have been deleted from Table 7.5-4, Area Radiation Monitors.
The NRC ' L was informed of these errors in the FSAR by a station employee.
. ! When informed, the licensee performed a technical review of Section 7.5.6, l Radioactivity Monitoring and identified a number of needed changes to this .section.
These changes were submitted to the NRC in FSAR Change 90-MP2-3 t
and were implemented on June 20, 1990, as part of the required annual update to the FSAR. The errors resulted from a modification performed in the 1980-82 time period.
' To evaluate ti,e effectiveness of the licensee's current program to main-tain the FSAR current, as required, three plant design changes reported , in the January 1 to December-31, 1988 annual report were reviewed to verify they were incorporated into the FSAR.
These design changes were DCM 2-020-87, Alternate Feedwater Flow Measurement; 2-027-87, Anticipated i; Transient Without Scram (ATWS); and 2-023-87, Diesel Air Start System L Modifications.,-These changes were verified to have been reflected in the '~ FSAR. Although an error appears to have been made during the-1980 - 82 period, the licensee currently has procedures in place to assure that , plant chonges are incorporated into the FSAR.. No-further action on this item is required.
, , I l-3.2 Incorrect lighting Panel' Breaker Tagging I The issue involved a lighting panel breaker red-tagged "open," but-identified.as " shut" by maintenance personnel, p Administrative Control Procedure, ACP-QA-206A, " Station Tagging", i requires, in'part, that operations personnel verify that all equipment p is in the proper position.. However, the control of nonessential light-ing-panel breakers during maintenance is not a' regulatory concern. The ., . discrepancy was identified by plant personnel to plant operations and r prompt corrective action was taken. 'It is indeterminate whether or not ' the ' tag was hung on the' shut breaker or if the breaker position was
changed after the tag was hung.. Since corrective action was taken, no' ' further action is'needed.
- p.
'. _ _ j !
[ !, j.. , ' . ^ p
3.3. Wrong _ Power Supplies !_dentified_In Plant Design Change Evaluation (PDCE) ! n ' W2-89-040' AND MF2;M-76 On August 8,1989, a licensee employee informed the NRC of certain errors j-in PDCE MP2-89-040. This PDCE was prepared to relocate a space heater and to relocate / eliminate several lighting fixtures to facilitate the addition of a hot machine shop in the auxiliary building.
Specifically, certain y electrical circuit breakers specified as providing power to several lighting fixtures were incorrect.
These issues were turned over to ! the licensee for resolution. On September 5, 1989, a licensee employee informed the NRC of certain errors in PDCE MP2-89-76.
Also, as a result c of a number of similar problems identified with PDCE's, the review and approval process for PDCE's was questioned.
The PDCE in question, was written to provide for the installation of duplex receptacles on the outside walls of the Millstone Radwaste Reduction Facility (MRRF).
'Specifically, it was noted that one duplex receptacle was shown as being powered from a wrong circuit.
The licensee prepared Automated Work Order (AWD) M2-89-08884 to investi-gate the drawing discrepancies which caused these errors and to determine the correct circuits. and lighting schedules in the area.
Also, AWD M2-89-08886 'was written to determine the correct electrical loads on the breakers: involved.
In addition, a Document Change Notice M2-89-040-2 .was issued to correct the PDCE which contained the errors.
To correct PDCE MP-2-89-76, a DCN was_ written to correct the error and to make the applicable drawing changes.
Following identification of this issue an as-built verification of all electrical circuits in the MRRF butiding was made. A number of similar problems were identified and appropriate drawing changes were made.
Based on discussions with personnel, it appears the errors in the PDCE's result from the use of incorrect record drawings and do not result from-a faulted ~PDCE process.
In_the past, these types of drawing errors have been corrected as they were identified.
This has not prevented similar { errors in PDCE's from occurring.
, The root cause of this and other similar errors is that lighting panel schedules and the associated drawings provided following plant construc- ,' tion contained errors.
To correct this deficiency, a Project Assignment-89-081, Lighting Panel Load Verification;and Drawing Update _has been i, * issued _in order to update all plant lighting drawings and panel schedules.
l This effort is scheduled to begin during 'the current-refueling outage.
As- ~ an interim measure, a memo was issued on August 14, 1990, to verify light- , ing panel circuits prior to performing planned; work.
These measures should correct the problem both short and long_ term.
No further action on this
- ~
matter is needed, , t<
. _ _ _ _ _ _ _. , ( g w
o '.
e
. hY
l
, 3.4 Reactor Coolant Pump (RCP) Motor Discrepancies i !- ! On, August 8, 1989, a licensee employee informed the NRC that during ' the performance of work on a spare RCP motor, four hoses were found L' unexpectedly to be disconnected.
This issue was turneo-over to the , i licensee for resolution.
> i This RCP motor hed been sent offsite to a General Electric (GE) facility ' . for overhaul and modification. After being returned to the site and while- ' '. being prepared for installation, four hoses were found unexpectedly to be ! ' disconnected.
Following the identification of the disconnected hoses, the + c licensee contacted GE to resolve the matter. GE had removed the hoses _! ' during the motor modification and the hoses were left disconnected when i the motor was shipped to Millstone.
The hoses were determined to be
+ the supply to the reverse rotation DP transmitter. With the aid of a- ! GE service representative, the hoses were reconnected to their correct l location. The motor was inspected for other anomalies, tested, and ! subsequently placed into service.
The motor has operated successfully
during this operating cycle.
, ," . . As indicated by the licensee, in order to prevent recurrence, future 6-purchase orders for RCP motors sent offsite for overhaul and modification a will be written to identify specific work which remains to be performed following return of the motor to.the site.
No further action on this item i is required.. - 4.0 Exit Meeting [ '- t Licensee management was informed of the scope and purpose of the-inspection
at'an entrance meeting-conducted on September 10, 1990.
l The findings of.the inspectors were discussed' periodically with licensee.
. [ representatives during-the course of the inspection,- An exit was-conducted on. September 14, 1990, at which time the findings of~the inspectors.were
presented.
. , At no time during the inspection did the inspectors provide written ! material to the licensee nor did-the licensee. indicate that areas covered . by-this' inspection contained proprietary information.
'; i
. ! , .-
, '
!1 _ _. _ - - _ - _ - _ _ _ _ _ _ _ - _ }}