IR 05000341/1990009
| ML20058L646 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/25/1990 |
| From: | Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20058L644 | List: |
| References | |
| 50-341-90-09, 50-341-90-9, IEB-90-002, IEB-90-2, IEIN-87-023, IEIN-87-23, NUDOCS 9008070368 | |
| Download: ML20058L646 (19) | |
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i U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-341/90009(DRP) , Docket No. 50-341 Operating License No. NPF-43 Licensee: Detroit Edison Company 2000 Second Avenue Detroit, MI 48226
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Facility Name: Fermi 2 Inspection At: Fermi Site, Newport, Michigan Inspection Conducted: May 14 through July 6, 1990 ' Inspectors: W. Rogers S. Stasek * F. Brush D. Butler J Gavula J. Smith J.'Jacobson * D. Schrum Approved By: fN R. W. DeFayette, Chief 7 Er[N# l
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Reactor Projects Section 2B_ Date *
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Inspection Summary ;
. Inspection on May 14 to July 6, 1990 (Report No. 50-341/90009(DRP)) , Areas Inspected: Action on previous inspection findings; operational safety; maintenance; surveillance; followup of events; LER followup; instrument tap
- failures followun; Ti11 item followup; Bulletin followup and Information Notice L followup, Results
- Operations adherence to administrative controls still reflected *
weaknesses as evidenced by an example of a violation in the surveillance functional area. However, adherence to the administrative controls for b Limiting Conditions for Operation has improved. The lack of job planning / ,
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l proper tool staging continued to cause delays in accomplishing maintenance activities. A previously unidentified weakness in the technical support function of material / component review, which caused the reactor core isolation cooling system to be degraded for a short period of time, was rectifie Reactor engineering technical support to operations was good. Some weaknesses, though, were noted in the clarity of and definitions in procedures i
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Pd.M' u. sed by reactob engineering.a e Howeverp these' weaknesses l didjoti ca' u se;" re,r' o.bM .s"i i*
* in' the implementation:of the program primarily due to well Mualified and 4 . . f V ji SArt 4 trained personne1Linithe' support positions. Two nonicited violations!were L f O m N $7%$ identified in ' Paragraphs.3,a. , and '3.d. ,and 5'(one violat f ori, tMexamplesp , 7 : N r iW ' O b 7one unresolved item was identified in Paragraph 4 and onCopen,(item vas " < t! - #,1 . M ,
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DETAILS-
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l Persons Contacted L, e Detroit Edison Company
< *P.. Anthony, Licensing "
c L *R. Bryer, . Supervisor, Material Engineering Group E *S. Catola, Vice President,' Nuclear Engineering and Services ' f
*G. Cranston, General Director, Nuclear Engineerin : *
P. Fessler, Superintendent, Technical Engineering
*D.' Gipson, Assistant Vice President / Plant Manager 4 *L. Goodman, Director, Licensing h *K.Howard, Supervisor,' Plant Systems 4
*A. Kowalczwk, Superintendent, Maintenance ,
,, R. May, Director, 1uclear Materials Management
~* R. McKeon, Superirstendent,.0perations
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* .1 * Miller, Director, Plant Safety , - . ,
G. Ohlemacher, Principal Engineer, Licensing _
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5 * Orser, Senior Vice President, Nuclear. Operations'
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9 ' J. Pendergast, Compliance Engineer
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J. Plona, Operations Engineer ,
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B. Sheffel, Program Manager, ISI, Technical Engineering '~i _J <
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~F.:Svetkovich, Superintendent, Operations
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R. Stafford, Director, NQA t ' l ., .
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! ' W, Tucker, Assistant to the Vice President .
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g *J. Walker, General Supervisor, Plant Engineering; s '
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U. S. Nuclear Regulatory Commission (" . . , ,
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- * Rogers Senior Resident Inspector #
[' # *S.L Stasek, Resident Inspector- ,
*M.LDapas, Operations Engineer, NRR 4, s
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F.cBrush, Reactor Inspector, DRP/TSS
. D. Butler, Reactor Inspector . , , J. Gavula, Reactor Inspector ,a , i ! J. Smith, Reactor Inspector i J. Jacobson, Reactor Inspector c D. Schrum, Project Inspector ' ,[ * Denotes those attending the exit meeting on July 6, 1990.
l- tThe inspectors also interviewed others of the licensee's' staff during ] . this inspectio , Action on Previous Inspection Findings (92701) (Closed) Open Item (341/88027-02(DRS)): Channel check acceptance criteria to be re-evaluated for practical limits to detect excessive channel drift. Main eteam line "A" temperature trip unit B21-N615A was found outside its as-found calibration tolerance by 8.85 degrees
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fahrenheit. The excessive drift of 821-N615A appeared to be an . r'
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isolated case. The NRC inspectors reviewed Surveillance Procedure NPP-24.000.02, "Shiftly, Daily, Weekly, and Situation Required
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3 Surveillance,". to determine if the' channel check acceptance criteria
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was adequate to detect a drifting temperature channel. The Li
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procedure's acceptance criteria indicated that all indicators ~must 4
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agree within'30 degrees Fahrenhei D The licensee perfornied a comprehensive review of past. readings of ' the main steam tunnel line temperatures. The review determined that the steam line temperatures deviated as much as'20 degrees c .,
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Fahrenheit between lines. The licensee modified procedure L
, . NPP-24.000.02 to notify instrument and control (I&C) personnel'if re the indicated temperature differential was greater than 25 degrees > ,!. " . Fahrenheit and to have I&C perform a calibration on the suspecti , ,, instrument (s). The original 30' degrees Fahrenheit channel check acceptance criteria was left as i '
The licensee's corrective ; J ,
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action was considered adequate by the inspectors and this item - is considered close , , L
.' (Closed) Open Item (341/89020-03(DRS)): Additional procedural ', ,t controls to ensure that valve E11-F027A is closed when using the .-
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, dedicated shutdown panel. The licensee added Step No: 28.d., ,
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" Manually close or verify closed E11-F027A, Division'I Torus Spray Valve," to procedure number NPP-20.000.18 (Revision 8), " Control
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of the Plant from the Oedicated Shutdown Panel." The licensee's; corrective action was considered adequate by the inspectors and- ' this item is considered close >
, , (Closed) Unresolved Item (341/89020-05(DRS)): Develop a better < , , method for determining tank level. The licensee's review of this "
item confirmed that the original design of the two standby gas treatment carbon dioxide system tanks included the existing '
, mechanical float' level indicators as an integral part of the tan ;
! The licensee noted that this was the manufacturer's typical design l for this size tank (1.25 ton) and met the requirements fo fire L protection service. However, the mechanical float indicalors could ! only be calibrated by draining the tank and opening the irispection j_ cove ' l' . I Following the 1989 NRC fire protection inspection, the licensee ' contacted the manufacturer and learned that the vendor had .
' : "." discontinued the mechanical float design for level indication and ;
y - currently utilizes a differential pressure desig The licensee-
- learned that a conversion kit was available to install the
% i differential' pressure type of level indication, which can be c'alibrated without draining the tank. Subsequently,1by letter dated ,
- September 25, 1989 (NRC-89-0183), the licensee committed to replace ,
the existing mechanical float indicators with the new differential, , pressure typ The new indicators were installed under EDP 1066 " This item is considered closed.
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f ', o , (Closed) Open Item'(341/89018-02(DRP)): Repainting the ""
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noninterruptible air syste The licensee completed painting both
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divisions gra ' (Closed) Open Item (341/86019-01(DRP)): Instrumentation lineup . 1 in walkdown order. ' The Utensee provided the inspector with m W ,' information that instrumantation valve lineups had been revised to place the instruments in walkdown orde , (Closed)'Open Itew (341/89018-09(0RP)): Elimination of the backup;
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manual scram breakers. The breakers were removed under EDP 10127, during-the first refueling outag (0 pen) Open Item'(341/89018-06(DRP)): Resolve emergency equipment
' ' service water flow discrepancies. The' implementation of engineering - - - , design package (EDP) 9487 to resolve this problem is scheduled for completion by the end'of July 199 '
y e h .- (0 pen) Open Item-(341/89030-0G(DRP)): Temporary lighting improvements for the drywell during outages. The design work for EDP 11044 is scheduled for. completion by August 7, 1990 and will 4 t( provide for permanent power feeds in the.drywell. Implementation
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of the EDP is scheduled to be accomplished at the beginning of the; .g second refueling outag f
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.L E' (Closed) Opn Item (341/88012-04(DRP)): Nuclear Shift' Supervisor ,
/ (NSS) Work Request Log. The. licensee initiated the w'ork.requesta 4 3
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(WR) log that includes' espies of all WRs that have received}NSS. g ,
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\_' N 3J approval to start work and is maintained in the NSS office < The \- , c, *
d inspector reviewed the WR log and verified consistency with the V, . ' packages currently approved by the NSS to work. Thiv item is' '
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considered close + i s t
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Instrument' tap;failureson4
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(0 pen) Open Item (341/90005-05(DRP)): , ,
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'J., ', , the residual heat removal system. Paragraph 10 of 1.nspection report; ,
341/90007 and Paragraph 11 of this report provide an update of '
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of.i ; licensee actions and t'9C review of this condition. di '
. % (Closed) Non-cited Violation (341/90007-05(DRP)): Documentationi. ..
E f inadequacies in cooldown records of April 14, 1990. Tne inspection' 4 effort documented in inspection report 341/90007 was sufficient to q close this ite '
. > , Operational Safety Verification: (71707)
The inspectors observed control room operations, reviewed' applicable logs and conducted discussions with control room operators during the period
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from May 14 to July 1,'1990. The inspectors verified the operability of
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selected emergency systems, reviewed tagout records and verified proper m i
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i return to. service of.affected components. Tours of the reactor building' - h and turbine building were conducted to observe plant equipment ,1 !
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conditions, including potential fire hazards,' fluid leaks, and excessive
- vibrations and to verify that maintenance requests had been initiated for, -
o: equipment. in need of maintenanc <3 3
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The inspectors, by' observation and. direct interview, verifi'e that'the '[ physical security plan was being implemented in accordance with the' ,
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station security pla ; ,
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; s t 1 The inspectors observed plant housekeeping / cleanliness conditions and '
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l' verified implementation of radiation protection controlst < During the ,
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i , , - inspection, the inspectors' walked down the accessible portions of the'
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s , following systems to' verify operability by comparing syste ,
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'.- plant drawings, as-built configuration or present valveli,m neup.lineup lists;with '; '
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Ji - observing equipment conditions that could degrade performance; and-
; '7 _ . , verified that instrumentation was properly valved, functioning, and ) % calibrate '
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* Standby Gas Treatment System - Divisions I and II '
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' Emergency Diesel Generator No. 11
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* High Pressure Coolant' Injection System These reviews and observations were conducted to verify that' fak:ility, '
operatio'ns were in.conformance with the requirements established unde ~ u e technical specifications, 10 CFR, and administrative procedure , , During 'a control' room walkdown on June 6,1990, the inspector noted
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:l a reactor core isolation cooling (RCIC) drain pot valve, E51-F025,- .
closed. The normal position is open. l.ater that day the I&C shop , replaced the solenoid on E51-F025. Following-a post installation review I&C personnel determined that the solenoid installed was;a *
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VAC instead of a VDC solenoid. Also, the original solenoid that had failed on June 6, 1990 was a VAC solenoid and the solenoid on the i E51-F026 valve, the. companion drain pot valve, was VAC. Adequate r ; compensatory measures were established by the operating authorit until VDC solenoids could be procured and installe ~j The licensee's review of this matter determined that the VAC solenoids were installe( 2n May 23, 1990 under the licensee's i environmental qualification (EQ) replacement program. Prior to implementation of the work request to replace the solenoid, planners o identified through a stock number in their computerized stockTsystem that the needed solenoids were in the warehouse and that this stock number correlated to the solenoid manufacturer's (ASCO) catalog 6 . number. .However, they noted that this stock number was not ! authorized by the material engineering group (MEG) for application for these two valves, E41-F025 and E41-F026. The only valve applications were approved for VAC solenoid valves. Therefore, '! the planners requested the MEG to review this stock number against !
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I these two' valve applications. MEG performed the review in, accordance with its administrative controls procedure, which consisted of a verification of catalog, number to stock number, and accepted the two new valve application However, to assure that the right ASC0' solenoid was used for a particular application required more than a' catalog check because the catalog number for a VDC and a VAC solenoid were exactly the sam The licensee's materials review program did not take that
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into account and the licensee's program for accepting a vendor on the approved vendor's list did not require a unique identification system for every component applicatio Installation of the wrong safety related solenoids on the RCIC system is considered a violation (341/90009-01(DRP)) of 10 CFR 50, Appendix'B, Criterion VIII, " Identification and Control of Materials, Parts, and Components." After reviewing this matter, the inspector determined that the criteria specified in 10 CFR 2, Appendix C, Section V.G were satisfied and therefore no' notice of violation will be issued. .The corrective actions included:
* Review of the VAC stock number applications to assure they were correct, * Disapproval of E51-F025 and E51-F026 for the VAC. stock numbe * Establishment of a new stock number for the VDC solenoid * Revision of the standard receipt inspection form for ASCO , '
solenoids to verify nameplate dat , i e Review of the material control system for all ASCO solenoids to s 4 - assure the proper solenoids were approved for each applicatio .
* * A change to administrative procedure MMP-PM1-01,)' Technical Review of Procurement Documents," to require at leact on '
functional parameter of a component be checked beyond the par ,
' number for safety related components and modification of the -
t , specific requirements for ASCO to include voltage, frequency, pressure, and medi ,
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, * Questioning of all other approved vendors (63' of 141) where ' -
this same situation was creditable to determine whether the-
' + same part number could have different applications.s Cne other' i - vendor, Okonite, indicated that this could occur'since a new + M tape product line may receive the number for an obsolete ' * ,
product-line after an unspecified period of time. The licensee , Q modified procedure MMP-PM1-01 to take this into account when i conducting material review ;
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* Training of most of the MEG personnel on this situation and I' r how it happened by the end of the inspection period, with two '
personnel to be trained in July when returning'from leav , The root cause of the violation was an. inadequacy-in the licensee's \
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approval process for the contr'l of material,' parts and_ components. . ; This inadequacy has been rectilied by the licensee's' corrective
, actions. This violation (341/90009-01(DRP)) is considered close ; The inspector reviewed Abnormal Lineup Sheet (ALS) 90-0489 which involved a tagout of the Division I Control Air Compressor, tThe ' > ,
inspector noted that a markup of a Functional Operating Sketch (FOS).. ., _ , that was included with the ALS, was not consistent with the full-size '
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Mylar drawing nor with the the actual tagging boundary ;When~ brought to the attention of'the tagging center NSO, the erroneous drawing was removed from the packag The NSO indicated that.the v ; reason for the inclusion of such a drawing in the package was i unknown and further, not neede .
. " During a walkdown of the Division II switchgear room, on May 26, '
3990, the inspector noted that' deficiency notice tag (DNT) 008198 ,
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dated November 3,i1989 was hung on MCC 72F-2A and indicated that th r flasher unit on the ammeter for the Division.II Battery Room Fan
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Coil Unit was not operating correctly. Further review found that work request (WR) 007C891113 was initiated to troubleshoot the-
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' problem. The WR was subsequently worked with no equipment problem found. The package was' cancelled on May 2, 1990, however, the DNTl 3 l had not been removed. The DNT was subsequently removed by th i
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, Reactor engineering support to the operating staff was reviewed by ,,
l e region based inspectors. Personnel were:well_ qualified for their j
support roles. Controls for inputr, to the process computer appeared " adequate with engineering design packages (EDPs) and/or administrative
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controls necessary for computer software change ' One error was noted in the implementation of these.contro s in that ,
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Procedure 53.000.04, " Change to Process Computer Data' Class Constants," L
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required an EDP to change any constant other than a select few listed
, - in Enclosure A of the procedure. Certain constants not identified ;
in Enclosure A were changed during refueling out' age 1 and during l 1 operating cycle iy This problem is considered an example of a violation of 10 CFR 50, '
' Appendix B, Criterion V, " Instructions, Procedures and Drawings," ; (341/90009-02A(DRP)). However, inspector review has determined that j a notice of violation is.not warranted because this matter meets the ^ criteria of 10 CFR 2, Appendix C, Part V.A.- Specifically, there have been no violations in this area in the last two years and the I problem itself was not safety significant since the changes;in the !
a constants were technically correc ' y
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Another weakness noted during the inspection was the vagueness!of some of the terms and phrases utilized in the reactor engineering
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procedure Two examples were " technically qualified member of'the , unit technical staff" and " operability of the process computer." ' The surveillance attributes of this support function are discussed in Paragraph 5 of this repor No other violations or deviations were identified in this are ,
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. Monthly Maintenance Observation (62703) '
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, Station maintenance. activities on safety-related systems and components 3
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listed below were observed to ascertain that they were conducted in
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L .accordance with approved procedures, regulatory guides and industry '
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f ' codes or standards and in conformance with technical specification . The following items were considered during this review: the limiting conditions for. operation were met while components or systems were' .; removed from service; approvals were obtained, prior to initiating the ;
.work; activities.were accomplished using approved procedures and were i
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inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified;
, radiological controls were. implemented; and fire prevention introls were
> implemented.- Work requests were reviewed to determine the status of outstanding jobs
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[ The following maintenance activities were observed: , t^ ..
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Y' * * The Standby Gas Treatment room cooler preventive maintenance (PM) C y . appeared to be inadequately organized in terms of paperwork preparation, < Fs %. S : staff preparation, and management oversight'of maintenance personne ,
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* * .The maintenance perscnnel were' prepar'ed to begin workion the PM and' <
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- ~ discovered that on6 af the procedures in the work package was not;i , i .(e the most recent revision. This caused a delay in the start'of work 1 , h , . , ,
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the stockroom when maintenance personnel tried to obtain' supplies.>.
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, Grease guns had b u n left out in the plant during pervious jobs co j l*
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C/ , were not available until they were located and returned to the - 4 + stockroom. In addition the specific grease ordered for'the w'ork '
' ** package had not been sent from the warehouse to the stockroom for .. '
performing this jo ,
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'3 task. In part this was due to a procedure that did not contain a . ~ ?,
complete material list. But in addition the maintenance staff had- *
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not thought out the details ~of.the PM and its material . ,1
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requirements. On two of the occasions the stockroom was closed, so J' maintenance personnel had to call stockroom personnel and wait for j the stockroom to be opened to obtain material ' ' ^'
* The' work' package did not contain important information to perform ,
the work. For example, it did not contain the fact whether the g" bearings are shielded or sealed. Greasing of the bearings was +" delayed pending a Preventive Maintenance Coordinator's determination ' whether to grease the bearings or not. In addi- n, work'was also S stopped to obtain assistance in finding the loc. on of set screws ' y so they could be staked, s ' The inspectors observed that the maintenance staff performing the-
' ~y ' ' * .
E' - work were deleting non-applicable portions of generic' procedures t,- y during the work process. This may be a violation of licensee 1 f '
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i, i procedures which require a supervisor's signature prior to deleting, J '; steps in the procedures. Pending additional review of'this issue, 3;< z
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S this is.an unresolved item (341/90009-03(DRP)). . .l , lt
' * 'Even though the overall PM was performed satisfactorily excessive . . '
s time and resources were expende The lack of organization of work ,' ?*b;
, . - appears to indicate that there may be insufficient attention to' t
<, . 4 . detail by management and/or a continuing lack of communication 4 between the support group '
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The inspectors also had concerns with another work activity'which' ' J 1 occurred in.the Standby Gas Treatment room. While observing the:I&C
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staff replace a transponder for control of a damper on safety-related
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_ equipment ~, the inspectors noted thtt the majority of the work was c' performed with channel lock pliers, instead of wrenches. This has a, m ;. , , , L potential for damaging fasteners. It appears to be another example of ,
,, .
( ' inadequate work practices and insufficient attention to detai ~'
,
Following completion of maintenance on the standby gas treatment room ,
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coolers, the inspectors verified that these systems had been properly * returned to service.. Other conditions that were affecting the mechanical h maintenance activities were'the licensee completing the final phase.of
. a commitment for. training. One third of the maintenance group is in a
'
l- ~ o 15 week training program. The other two thirds of the maintenance ' personnel had already completed this' training. This training should contribute to the long term improvement in the quality of the mechanical 1
> maintenance work. Also, the inspectors were informed by a maintenance ;
supervisor.that many of the mechanical maintenance personnel are near to their maximum yearly radiation exposure, thus limiting staff decisions ,' ,' for who works in high radiation areas. Thus some of the most qualified ' ' staff can not be selected to perform important maintenance wor '
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No. violations or deviations were identified in this area. ' ? J ,
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, ~ Monthly Surveillance Observation (61726) , ,1 !
The inspectors observed average power range monitor calibration testing, j
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NPP 54.000.06, "APRM Calibration," required by Technical Specifications' .
'
3- and verified that: testing was performed in accordance with adequate ' , procedures, test. instrumentation was calibrated, limiting conditions for?-
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operation were met, removal and restoration'of the affected components
*
were accomplished, test results conformed with Technical Specifications' and' procedure requirements and were reviewed by personnel other than the o individual directing the test, and any deficiencies. identified during the testing were properly reviewed and resolved by appropriate management'
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. personne .
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F The inspectors also witnessed portions of the following test' activities:
, * 24.204;01 Division 1 LPCI and Suppression Pool Coeling/ Spray Pump and Valve Operability Test * 24.'307.17 Emergency Diesel Generator No. 14 Start and Load , Test * 44.010/166 Flow Unit C Calibration ,
l C Regarding 24.307.17, the inspector noted that Heise type temporary pressure gauges had been installed on the fuel oil transfer pumps associated piping. Further review revealed the gauges had been connocted . the previous evening in preparation for the surveillance, NPP-24.307.37
"DDSW and DFOT Pump and Valve Operability Test - EDG No. 14", which was ' to be run subsequent to HDP 24.307.17. This installation was done out-of-sequence to the procedure, in anticipation of starting the tes However, the test was postponed until the following morning due to the delayed completion of ^ EDG No. 11 maintenance outag When the operating authorit, s apprised of'the situation, Deviation !
- Event Report (DER) 90-0340 ww initiated and a review was conducted to- :
i determine root cause. Subsequently, the licensee determined that unprecise work direction was given to a shift engineering technician '
' (SET) by his supervisor. The SET was directed to install the Heise gauges in preparation for the surveillance.. However, ensuring compliance ,
with (or performing the installation with) the approved procedure was l-
, ;( not specified (although the SET was aware the gauges were placed per a -
surveillance procedure). .I 4 a' , l- . . c6 c
In response, the licensee counselled the individuals involved, and * ?/
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/ ,- 'provided further guidance to the operating shifts on management's? 1 1 ' i expectation that surveillance procedures were to be sequentiall =
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i followed. This problem is considered an example of a. violation of , 1 ,+ c, s= 10 CFR 50,' Appendix B, Criterion V, " Instructions, Procedures and >
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However, inspector review ha * ' Drawings," (341/90009-02B(DRP)). + lr , s determined that a notice of violation is not warranted because this' [ .
' . 1 matter meets the criteria of 10 CFR 2, Appendix C, Part V.A.
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Another problem was noted with this surveillance activity'in that~ *
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subsequent.to completion of Surveillance Test NPP-24.307.17, the '
:. inspector reviewed the completed work package PM W843900126.and noted the post maintenance testing (PMT) form was not signed on by the NSS
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, documenting authorization for test performance. When questioned on this
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matter, the NSS indicated that the specified PMT involved performing the routine periodic surveillance that was currently due and the approval , for the' testing:to start was documented on the Surveillance Performance " i4- Form (SPF). Although the surveillance test was performed under the "
, ;, cognizance of and with approval of the NSS, a. portion of the licensee's - , ' administrative controls was not adhered to in that the PMT was not signed ?
aa by the NS .. . Further discussion with the operatinge'ngineer resulted in a night order i heing issued to all operating shifts, impressing the need for ensuring y, fud autber12ation of testing under both the surveillance program as well 1, . as th's p'ost maintenance testing program (as required).
b A record / technical adequacy review of completed surveillance tests i associated with reactor engineering was performed. The results were:
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* The computer generated power distribution reports P-1 and OD-6 were b examined at 99.9 percent rated power and all thermal limits were determined to be within their prescribed limit =t
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* The backup core limit evaluation (BUCLE) was adequate for determining core thermal limits when the process computer is unavailabl t .
i
* The local power range monitor calibration through process computer determination was satisfactory, e * The procedure for manual core thermal power evaluation wastadequat '
A record review of selected other complete surveillance procedures-was -
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performed. The review was to determine that the test was accomplished 4
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within the required Technical Specification time interval, procedural i, - steps were properly initiated, the procedure _ acceptance criteria were met, independent verifications were accomplished by people:.other than' e
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those performing the test, and the tests were signed in and out of the
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control room surveillance log book. The surveillance tests' reviewed ,
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were:
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* 44.120.001 Division I Channel Calibration' . ,
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* 44.030,257 ECCS-Reactor, Vessel Water Level (Levels 1, 2,'and 8), ;
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, lm e Div. I, Channel C Calibration / Functional A '
No other violations or deviations were identified in this are . a ;
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Followup'of Events ll'
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e, 'During the inspection period, the licensee experienced leveral' events, ) ' '
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4 s one'of which required prompt notification of the NRC pursuantJto ; i '
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1 '"( 10 CFR 50.72. The inspectors pursued the events onsite with licensee' ,
"
s and/or other NRC officials.- In each case, the inspectors verified' -
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^' that the notificatio'n was correct and timely, if ' appropriate; that the . -
licensee ~ was taking prompt and appropriate actions, that activities were
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L 3: , "y N.$ C conducted within regulatory requirements and that corrective actions would prevent future recurrence. The specific events are as follows:"
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, Identification of a licensed operator unfit for dut Ondune(12,l C 4
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1990a'seniorlicensedoperatorwasscreeneda unfit for dut This 'was confirmed by additional testing. . 'Thc = d '
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t , individual was removed from duty and the licensee is preparing a" .
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medical disability letter for the NR ,
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'% ~ Sampling of emergency diesel generator (EDG) fuel. oil tank a routine- fuel oil. sampling surveillance on April 27, 1990,.the During ;;
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.~ y me first oil sample taken on EDG 11 indicated higher than. allowed *
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particulate contaminates. . Subsequently; chemistry personnel tooko
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four additional samples. To second sample was satisfactory. The .
's. ', third was not. Following the third sample operations personnel were ;
contacted and the EDG declared i - erable. The fuel oil in the ., a ; tank'was recirculated through filt'ers. Thelast:two samples were ' satisfactory and the EDG. declared operable, o The inspector pursued the sampling methodology with the chemistry ' ,L I supervisor.. The inspector ascertained that the sampling technique,
, was appropriate, implemented properly, and'the sampling program - '
did not require declaration of the EDG inoperable until the first' unsatisfactory sample was" confirmed Ly another. sample. The inspector expressed concern' that this was a nonconservative approach and that the EDG should be' considered inoperable following the first unsatisfactory sample. This had already been independently ' " - '
, determined by the Operations Superintendent and.he had directed chemistry to discontinue the practice. Subsequently, applicable ,,
procedures were changed to reflect the new philosoph } c/ .Readtor Recirculation (RR) pump speed mismatch. On May9, 1990, during a= unit power increase, RR pump speeds;were intentionally , mismatched in attempts to avoid vibration problems associated with p , the B pu?p at the 70-76 percent speed region. Technical Specifications require less than'a 5 perced mismatch between the A two RR pumps at greater than 70 percent power. The operators were fi, . ~ aware of the requirements and maintained RR pump speeds within t% '&
,1 ; 5 percent band as indicated on the speed controllers. !However,Ja \
g . CRIS dot had previously been placed on the-A pump. speed controller , T indicating that actual pump speed was approximately '2:5 percent less -
' , than indicated. During the time when the pumps were-operating in a mismatched condition, the significance of the CRIS dot was noti e * ' ' >
recognized and resulted in the operators unknowingly operating theT ,
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recirculation pumps with a larger mismatch than intended. This "
; condition existed-until discovered by the STA approximately four hours later. . Initial followup by the operations department * '
concluded that the pumps had been operated outside the Technical '
[. - Specification requirements. However .further review determined .
#
that the B speed controller.was not indicatirg a full 2.5 percent iJ deviation from actual speed but, rather, was only about' half tha H
'Using the new' values , the licensee found that Technical ^ , Specifications had not been violate ,
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Although the Technical Specifications were not violated, the :- T operating authority still considered the oversight sip ificant in? &"
. that a question of operator cognizance of equipment s.atus was ,
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o raised Deviation Event Report (DER) 90-0342 was initiated on the , r.atter, and an operations event critique.(No. 90-010).was prepared' t - specifying the lessons learned from the event. The critique was included as part of required operator reading and the associated ' procedures were evaluated for potential changes to clarify the limitations:on RR pump mismatch. Additionally, nuclear training was evaluating inclusion of the event into the operator training s y progra ,,
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No violations or deviations were identified in this are '
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m Licensee Event Report Followup (92700) Through direct observations, discussions with licensee personnel, and ' * S i review of recordt,'the.following event reports were reviewed to determine
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P that reportability requirements were fulfilled, immediate corrective
. ' action was accomplished, and corrective action to= prevent recurrence had +
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,,- been accomplished in accordance with technical specifications. C e a
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(Closed)LLER 89-023, Loss of Power to Division 1 Due to Personnel
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#' < - Error by Relay Division, '; ., s ",'j ^ ,. , .
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i . p ,\ m i; b .' (Closed) LER 87-050. Unidentified Feedwater Piping,Desi'gn '. .
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y a Temperature Limit Exceeded, i L\ - ,
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c .i (Closed) LER 88' 015, Division I Solations Experienced.During
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.! K -I Implementation of a Design Chan P.- 0
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, (Closed) LER 89-012 and Revision 1, Inadequate Technical ^, Review of.' ~
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a Temporary Procedure Change Causes Division 1 Actuatioi)?
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e; (Closed) LER 89-022 and Revision 1, Reactor Protection System ' *
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d ' "n Actuations and Reactor Building Heating Ventilation and Air - 3 - Conditioning Isolation Which Occurred During the Installation of
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J' EDP 1012 t i-y 't: i,
+ $ 5
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ji:c - f, -(Closed) LER 88-034, Isolation of Reactor Water Cleanup Syst'em due T : to Suspected Relay failur , V@ffC4 _
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(0 pen) LER 88-034 Revision 1, Isolation of Reactor Water Cleanup'
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2 System.due to Suspected Relay Failure. In this LER update, the > Q; 3 l licensee documented the completion oi laboratory analyses confirming-D ' the presence of silver sulfide contamination. The licensee,is' * iS ' pursuing the source of the contamination with the relay supplier, ti,. General Electric. Also, subsequent to LER revision issuance, thel - l: ' t licensee decided to replace all the unshielded 120 CR relays in the i
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o relay room during the next refueling outage. Following discussion '
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with the_ inspector, the licensee will provide another update to the' i'+
'LER=when the review for the contaminates is complete and the full ' 7
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; scope of relay changeouts is determine ,l
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p , ,(Open) LER 90-003, Reactor Scram with Closure of'the Inboard Main
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H c Steam Isolation Valves. On May 8, 1990 the. licensee submitted F- LER 90-003.on the scram. The inspector reviewed the LER and noted '
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', some' discrepancies / omissions which were: l
+ * The first paragraph on page 3 indicated that operators entered ,
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"7 into the E0Ps due to SRV actuation at 0209, whereas, the E0Ps were entered into at 0202 with the closure of-the,MSIV >
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'*- The corrective actions section on page 5 stated that the
, 1 , leakage on indexer B was repaired during'the forced outage,
' ^ following the scram. This is not true, Work request 0040900419 was initiated on April 17, 1990 to repair the indexer. The work request is coded for cold shutdown and-
, .has yet to be complete '
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' * -The LER did not discuss the failure to, vent the RCIC discharge ,
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,1 .line to the CST or the undersized motor associated with-valve -
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E41F011. Upon reading the LER, one concludes that the reason' , 7 ;' for E41F011 failure was exclusively an equipment failure and i had nothing to do with. system design or operator trainin "l
* The additional training, provided to a11~ operating shifts, 5 <
was not discussed in the corrective action section of the LE %' Furthermore, there was no mention that when operators. reset
; .the scram the first two times it was not.in.accordance with l . *
previous training; that exclusive use of SRV A was inconsistents , n with operator training; and that operators entered the' E0P on a wrong entry conditio : o There was no discussion of the 62,000 gallon water spill in J ' the turbine buildin , } .. f l
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% , , * The cause of the event as stated in the LER was: "The< amount k T of leakage through the TIP indexer boxes had not been 4 1 m ,
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o ' quantified previously arid there is no acceptance criteria for
! ' *
1eakage established for this system." The corrective actions f section did not discuss the corrective action for. establishing
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laakage testing and a leakage acceptance criteria for the'
., TIP system /drywell pneumatics syste , ,
These observations were discussed with licensing personnel who+. L, ,
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stated that these matters would be/taken into consideration fo ' q h' the possible issuance of a LER revisio is 'j ,
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$ (Closed) LER 89033, Actuation of Emergency Equipment Cooling Water 1 due to Personnel Erro The only action left to be acted upon is
{1[' ; a recommended design change, PDC 11249, to cabinets H11P627 an H11P626 to improve accessibility. There have been two other,
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unplcnned actuations in these panels in the last two year E , Licensee disposition of PDC 11249 is considered an open item , , l c. - ,
.(341/90009-04(DRP)). ,
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[ '
' (Closed) LER 90-0004 Resident Heat Removal System Small Bore ,
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', Connection. Based on the reviews conducted by Regional Inspectors
~, v , as documented in Paragraph 11, this item is considered close .
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No violations or deviations were~ identified in this are l Followup of TMI Action Items (NUREG-0737) '
.During the inspection period, the following THI action item was reviewed:
(Closed) 1.D 2 Plant' Safety Parameter Display Console:
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L In a memorandum, dated January 9, 1986, the Director of the Office of Inspection and *
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L,' Enforcement informed the Region III Regional Administrator that review of L o the' safety parameter display system (SPDS) would be accomplished by NRR ,!
. audits of SPDS. Closure would be contingent upon those audits. In' M'irch j a
1987 a NRR audit was performed and the results of the audit were y documented in a letter, dated July 29, 1987, to the license SPDS was- J L not considered as meeting the requirements of NUREG-0737 in the lette ', I ' Fo110 wing 1 numerous acetings and corrective action,'the licensee j documented completion of all commitments on SPDS to the NRC in a letter
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dated January 11, 1989. A followup SPDS/DCRDR audit was accomplished by i < NRR in 1989. Subsequently, NRR accepted SPDS in a letter dated March 28, i 1990 to the licensee. In addition the inspector verified that an L operating procedure existed for SPDS and randomly selected a licensed- , ..
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operator to interview / demonstrate use of the system. All aspects of the ! q interview / demonstration were' found satisfactor #
- IE Bulletin Followup >
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(Closed) Bulletin 90-02, Loss of Thermal Margin Caused by Channel Box _
Bo The inspector reviewed the licensee's revised procedure, , 1 NPP-57.000.08, " Core Component Management Guidelines," addressing bowi.ng 2 l and considered them adequat ,
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(Closed) Information Notice (IN) 87-023, Loss of Decay Heat Removal
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Durine Low' Reactor Coulant level Operation. The licensee reviewed-the C IN undar DER 90-0194 and determined that the present management program 1 ,.
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encomptssed all of the applicable attributes necessary to preclude this 1 even *
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11.- Fatigue Induced Failu m of RHR Piping 4h
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j 4;' - During preoperational testing,of'the RHR System in 1982, piping and pump _ vibration concerns were noted.. As-a result of these concerns,!the ; 3x s
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licensee, the pump manufacturer (Byron-Jackson) and consultar!ts performed - l 5 , field vibration' testin Further evaluation and testing led to eventual f ; disassembly of the RHR pumps. General Electric and Byron-Jackson: < ' '
'
i concluded that excessive bearing wear had resulted from1 operation,at r >
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J ' i minimu.n flow conditions. Operation of the pump at minimumiflow apparent 1y'" b
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L %. produces shaft. side.Ioads, pressure pulsations and internal, recirculation: -S
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These ._"'l which results _in premature pump bearing degradation.e unique in' thatc,there a
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9' same ,
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4- size /NPSH/ flow pump' design supplied by the same vendo '.,,
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s 4 j The NRC inspector reviewed the applicable test reports and corrective ' p3 * actions taken by the licensee' to improve the- RHR pump performanc > > Corrective actions included pump internal modifications, installation of multiple breakdown orifices to reduce flow induced vibration, and bracing * , . , of,the'RHR pump suction.'and discharge nozzle to pipe interfac These 1 actions, though not totally effective, demonstrate substantial effort on
>
the part of the licensee to reduce pump vibratio Pipina vibration, evaluated under a separate program, is described in
, the. Updated Final Safety Analysis Report. Data accomulated from piping
. greater than four inches in diameter was analyzed and the results considered acceptable prior to low power testing in 1985. The data for l piping branches less than one inch in diameter was compared to acceptance
, criteria based on simple beam mathematical'models and was.also fou % ,
acceptabl t 1-
' ,, Beginning in December of 1986, vibration related failures in the RHR pump rooms were experienced.~ A RHR system vibration monitoring and analysis ' .-'
program was establishv in October, 1987 as a result of these failure This program was estc. iished to evaluate those components not covered.by the original vibration evaluation previously discussed. During the
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evaluation period several . vibration related failures of RHR pipe taps'
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and seal water cooler inlet and outlet connections occurre The results of this vibration monitoring program were documented in a' f preliminary reportin March 1990. This report indicated that several l' ' small bore branch connections were experiencing vibratory stress levels
in excess of the 10 Ksi acceptance level specified in ANSI OM-3.- ,
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y, y Modifications to these connection's (EDP-11441-1) were performed to reduce '
.. M N* ,, *$( the stress levels. The following supporting calculation / evaluation was '%
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reviewed by the NRC inspector: ,
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* Calculation No. 5135. Revision 0, May 19, 1990, " Qualification'of'
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4]. Small Piping Near RHR Pumps for-Steady State Vibratfor."- ' i
W, This calculation documents the initial results'of the stress'
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evaluations, using the original vibration data. It was'the '
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J G ?~' preliminary results from this calculation that prompted the April t
..m discussions and meetings between the licensee and the NRC. lThe NRC
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b , inspector noted that the methodology used in the calculation was + considered to be conservative. No discrepancies were notedt. This. ,
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calculation is. currently being revised .to incorporate the> results of
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a the modified configuration (EDP-11441-1) with new vibration dat m L y ' '. , , , w
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.c * "RHR Vibration Data Evaluation," May 31, 1990'and-June 14, 199 . G y
This preliminary document evaluated the RKR process line small bore - connections which were modified under EDP-11441-1. Using new . , 'i* '"X g vibration data,' obtained after the modifications were completed,, J stresses were conservatively calculated for~each connection. In all-cases, the pipe stress at the toe of the fillet weld, where. fatigue q ", j i cracking typically occurs, was less than the maximum acceptable s'
,
value of 7690 PSI. In addition to this theoretical work, e l discussions with the licensee indicated that the highest' stressed t @ l connection would be instrumented with strain gages to confirm the %y L conservatisms in the analyses. This strain gage work.was 4 , h y preliminarily cheduled for Fall 199 Y , m .. e :, As documented in the Deco letter, dated May 11, 1990 from W. S. Orser'" ; to the NRC, a commitment was made to visually inspect other RHR
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. vents, drains and test connections which were not included-in the 'h '
recent' efforts. This basically encompassed RHR connections which'
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f" were not in the pump rooms or in the drywell. For connections with "1 :% A " susceptible" configurations vibration data will be taken and dye '
%. i penetrant examination will be performe t '
. ? ) ' The NRC inspector reviewed the documentation associated with the y~i4
,. assessment of these additional connections. A' total- of 69 . > "e additional test / vent / drain lines, instrument connections,. relief J 'l^
o valve or keep fill system connectior.s were identified on both
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, divisions of RHR. Of these, approximately 11 connections were , m identified as having " susceptible" configurations. None;of3 these
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', " susceptible" connections were considered to be of any immediate , <
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assessment, ut w'
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4a + 4 Also documented, in the May 11 letter, was a DECO commitment to 4 review susceptible test vent and' drain connections >on~the prinary
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fR, lO - -r system pressure boundary piping. The NRC inspector,also reviewed '
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the documentation associated with this assessment / Approxi.nately)0 s .
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co'nn$ctlon1 configurations were identified 'by' the licensee. Most ofL
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these were standard configurations and not-considertd 'l susceptible."
t Of the non-standard configurations', most weretadjaceat to structurali f+ , Lan'chors and were excluded-due'.to theElack of. header vibratio c+ However, two test connections on the feedwater system _were-
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%~ considered " susceptible" and it was the" recommendation of DECO-
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; engineering to " butter" these welds during,the next outage.: The- i, '
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c 'NRC inspector concurred with this assessmen .
. In addition to.the number of 3ocket weld failures attributed to-Es
vibration, several were found to exhibit porosity defects'.s In an'. ) n y; - effort?to address the issue of weld quality? the licensee organized 4
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4 3a task force'to~ review socket' weld failure data for an-approximate'
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y , t * three This three = year period-(January 1987sto,May f 1990)- includes-year perio .
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s ,j' , the early stages of start up testing and power ,
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ascension 1through sustained commercial operation. Two, major events o 4 1?
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'(Main Steam and RHR failures) have already'been' addressed and werei H'-
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:not included in this surve .
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s Ihe survey identified 7 socket wela ioilures due'toYibration and 4
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ffailures, due to weld porosity, during the three' yeah periodr .Thei ' e ;" ', ,%Q - f: , failures occurred in 7 different-systems with different welders l ',1 >'
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identified for each. Of those failures identified,sno' safety-related
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systems were involved and there was a good distribution;o'ver tim .FromLthis data,- one cannot conclude any commonality which;would
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J ' question the overall quali_ty of socket weld g g i .
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M ' Additionally, metallurgical analysis was performed on four welds, *
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.three of which were removed as part of the RHR modifications and
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'one of which was removed-due to a1 pin hole leak. The'NRC inspectori '
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; reviewed both of tN survey results and' metallurgical analyuis and found no evidence of a generic welding deficienc m".
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< 3, ' 12. 'Open Items , . .d .y Open items =are matters which have been discussed with the licensee, whic > will be reviewed further by the inspector, and which . involve some action m" ^,. * , _ on1thelpart of the NRC or licensee orz,beth'. An open item ' disclosed iduring' the . inspection is discussed ,in Paragraph , ,
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' l13. Exit Interview (30703)
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6 The-incp'ectors met with licensee representatives,(denoted in Paragraph 1) '
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on Julyt6,1990, and informally throughout the inspection period ~and
; summarized _the scope and findings of the inspection activities. The "
inspectors':also discussed the likely inf ormational content of. the
, inspection report with: regard ~to' documents or processes reviewed by the
' inspectors,during the inspection. The licens9e did not identify any such The licensee acknowledged tha
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documents / processes as proprietar e
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findings of>the inspectio ,
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