ML18058B861
| ML18058B861 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/28/1993 |
| From: | Shafer W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18058B862 | List: |
| References | |
| NUDOCS 9306080109 | |
| Download: ML18058B861 (34) | |
See also: IR 05000255/1993008
Text
Docket No. 50-255
Consumers Power Company
ATTN:
Gerald B. Slade
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION Ill
799 ROOSEVELT ROAD
GLEN ELLYN, ILLINOIS 60137
General Manager
.
Palisades Nuclear Generating Plant
27780 Blue Star Me~orial Highway
Covert, MI
49043-9530
- Dear Mr. Slade:
' .
This refers to the inspection conducted by inspection personnel from this
office and the Office of Nu~lear Reactor Regulation on March ~3 through
May 17, 1993.
The inspection incltided a review of authorized activities
for the Pali~ades Nticlear Generating Facility.
At the conclusi-0n of the
tnspection, the findings *we~e discussed with those members of your staff
identified in the enclosed report.
Areas examined during the inspection are identified in the report.
Within
these areas, ~he inspection consisted of selective examinations of procedures
and representative records, interviews wit~ p~rsonn~l, and observ~tion of *
actiyities in progress.
Items discussed at a routine management meeting held
at the site on May 17, 1993, are also summarized.
The results of .this inspection showed that certain of your activities appeared
to be in violation of NRC requirement~, as specified in the enclosed Notice of
Violation.
The results also showed that management involvemen~during infrequently
performed a~tivities w~s a .notable strength.
However, there appeared to be an increase in th~ number of-operator errors.
These are summarized in paragraph 3e of the enclosed Inspection Report.
One*
unresolved item was identified for this issue.-
Both the violation and the
unresolved i tern are of concern because they suggest your inability to *
effectively resolve the increasing number of operator errors.
You are required to respond to this letter and should follow the instructions
specified in the enclosed Notice when preparing your* response.
In* your
response, you should document the speciffc actions taken and any additional
actidns you pl~n to prevent recurrence.
You are also requested to provide a
written response providing your evaluation and corrective actions for the
unresolved item addressed in paragraph 3e.
The response should be submitted
with the reply to the enclosed Notice.
After reviewing your response to the
Notice and the unresolved item, the NRC will determine if ~dditional
enforcement action is necessary.
9306v80109 930528
ADOCK 05000255
G
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Consumers .Power Company
2
Additionally, during th~s inspection, certain of you~ activities, pertainin~
to Eme*rgency Diesel Generator 1-1 inoperabil ity (pa'ragraph 7d), appeared to be
in violation of NRC requirements:
However, as described in the enclosed
. inspection report, yo~ identified this violation.
Thereforei the ~iolation
will not be subjett to enforcement action because your efforts in identifying
and ccirrecting the violation meet the criteria specified in Secti~n VII.B of
the "General Statement of Poi icy and Procedure for NRC Enforcement Actions,"
(Enforc~ment Policy, 10 CFR Part 2, Appendix C)'.
The responses di re.cted by this 1 etter and the enclosed Notice are not subject
to the clearance p_rocedures of the Office of Management and Budget as requ.i red
by the Paperwork Reduction Act of 1980, PL 96~511.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of
this letter, the enclosures, and your responses to this letter will be placed
i~ the NRC Public Document Room.
We will gladly discuss any questions you have concerning this inspecti6n.-
Enclosures:
2. Inspection Report
No. 50-255/93008(DRP)
cc w/enclosure:
David P. Hoffman, Vice President
_Nutl~ar Operations
OC/LFDCB
Resident Inspector, Rill
James R .. Padgett, Michigan Public
- Service Commission
Michigan Department of
Public Health
Palisades, LPM,
N~R
SRI, Big Rock Point
- Sincerely,
--f-?=:d-i ,_, .. . .. i
-..;.._0-.. .. _ j,..-f.* \\:__; ._:, \\,v~-:'---
-
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.
.
..
_s::.-~y-:::;-"W.* D. Sh-a.fer, Chief
- -
Reactor Projects Branch 2
MAY 2 8 i993.
- .
Consumers* Power Company
2
-Additionally, during this inspection, certain of your activities, pertaining
to Emergency Diesel Generator 1-1 inoperability (paragraph 7d), appeared to be
in violation of NRC requirements.
However, as described in the enclosed
inspection report, you identified this violation. Therefore, the violation
wi 11 not be subject to: ,enforcement action because your efforts in i dent ifyi ng
and correcting the Violation meet the criteria specified in Section VII.B of
the "General Statement of Policy and Procedure for NRC Enforcement Actions,"
-(Enforcement Policy, 10 CFR Part 2, Appendix C).
The responses directed by this letter and the enclosed Notice are not subject
to the cl~arance procedure~ of the Office of Management and Budget as required .
. by the Paperwork Red yd ion Act of 1980, PL 96-511.
In accordance with 10 CFR 2.790 of the Cbmmissfon's regulations, a copy of
thi~ letter, the enclosures, and your responses to this letter will be placed
.in the NRC Public Document Room.
We will gladly discuss any questions you have c6ncerning this in~pection.
Enclosures.:
2. Inspection Report
No. 50-255/93008(DRP)
c:c w/enclosure:
David P. Hoffman, Vice President
Nuclear Operations *
OC/LFDCB
Resident* Inspector, Rill
James R~ Padg~tt, Michigan Public
- Service Cammi ss ion
Michigan D~partm~nt of-
Publ ic Health
. - Palisades, LPM, NRR
SRI, Big Rock Poirit
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\\i.1\\~:>
bson ~
ra. 10a(5)
a d 10c(2))
Sincerely,
RII I
Rill
RII I
Jo:lis*n
-
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. ~rok
Consumers Power Company
Palisades Nuclear Generating Pldnt
Docket No. 50-255
License No. DPR-20
During an NRC inspection conducted on March 23 through May 17, 1993, a
violation of NRC requirements was idehtified.
In aciordance with the "General
Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2,
App~ndix C (1993), the violat~on is listed b~low:
Technical Specification 6.8.1.b requires implementation of procedures for fuel
handling activities. Standard Operating Procedure (SOP) 28, "Fuel handling
System," implements this requirement.
Attachment 5 of SOP _28 requires
permission of the shift supervisor prior to use of the spent fuel pool*
handling machine "Override Key Switch."
Contrary.to the above, on March 21, 1993, an operatcr used the spent fuel pool
handling machi~e "Override Key Switch" without permi~sion of the Shift
Supervisor.
A machine malfunction resulted in whic~ the main cable became
partially unwrapped and an attached spent fuel- assen-]ly dropped about .six*
inches.
This is a Severity Level IV violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, Consumers Power Company is hereby
required to submit *a written statement or explanation to th*e U.S. Nuclear
Regulatory Commi:ssion, ATTN:
Document Control Desk, *Wash.ington,
D~C. 20555
with a cbpy to the Regional Administ~ator, .Region III, 799 Roosevelt Road,
Glen Ellyn, Illinois, 60137, ~nd a ccipy to the NRC Resident Inspector at the
Palisades Nuclea~ Generating Plant within 30 days of th~ date of th~ letter
transmitting this Notice of Violation (Notice):
This reply should be clearly
marked as a "Reply to a Notice of'Violation" and should include fcir each
violation:
(1) the reason for the violation, o~, if contested, the basis for
disputing the violation, (2) the corrective steps that have been taken and the
results achieved, (3) the corrective steps that will be taken to avoid further
violations, and (4) the date when full compliance will .be achieved.
If an
adequate reply is not received within the time specified in this Notice, an
- order or a demand for information may be issued as to why the license ~hould
not be modified, suspended, or revoked, or why such other action a~ may be
proper should not be taken.
Where good cause is shown, consideration will be
given to ~xtending the respon~e time .
Dated atj-len Ellyn, Illinois
this A8' ""day of May 1993
9306080112 930528
ADOCK 05000255
G
a er Chief
rejects Branch 2
...
U. S*. NUCLEAR REGULATORY COMMISSION
REGION II I
Report No. 50-255/93008(DRP)
Docket No. 50-255
LiGensee: Consumers Power Company
212 West Michigan Avenue*
Jackson, MI
49201
Facility Name: *Palisades Nuclear Ge~eratin; Plant
Inspettio~ At:
Palisades*Site, Covert,
~~
Inspection Conducted: March 23 througn Mc~*
~ 7, *:. 993
Inspectors:
J. K: Heller
0. G. Passehl
W. 0 .. Shafer
Appr.oved. By:
- Inspection Summary
J.
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J.
A. Gavuia
A.
W. Ue.l SC~,
M.
F: Sch a~ <e,..
H.
E.
License No. DPR-20
Hsia
P::'.*ker
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Inspection from March 23 through May i7,
19~3
(Report tio.
50-255/93008(D~P))
Are~s -Inspected:
R6utin~, un~hnounced inspection by inspe~tion personnel from.
this office and the Office of Nuclear Reactor Regulati-0n of a~tions on
previously i~entified ~terns, operational safety verification, maintenance,
surveillince, ~ngiri~ering and technical sup~ort, reportable events; spent fuel
handling machine malfunction, dry cask storage operations, loading of the
first and second dry cask, and a quarterly canagement ~eeting.
No ~afety
Issues ~anagement System (SIMS) items ~ere reviewed.
Results:
One violation was identified involving failure. to *follow procedures
during fuel moves (paragraph 8).
One non-cited violation (NCV) was identified
pertaining to dies.el generator inoperabil it/ (paragraph 7d). * One unresolved
item was identified pertaining to an increasing trend in the number of
operator errors (paragraph 3e).
Jhe strengths, weaknesses, and inspection. fcllowup items are discussed in
paragraph 1, "Management Interview."
9306080115 930528
ADOCK 05000255
G
DETAILS
l.
Management Interview (71707)
The inspectors met with licensee representatives (denoted in paragraph
10) on May 24, 1993, to discuss the scope and findings of the
inspection.
In addition, the likely informational content of the
Inspection Report with regard to documents or processes reviewed by the
inspectors during the inspection ~as also dis~ussed. The licensee did *
not identify any such documents or proces$es as proprietary.
Highlights of the.exit intervie~ are discussed below:
a.
Strengths_ noted:*
1)
. 2)
3)
The administrative programs for minimizing the rt~k to
shutdown cooling during reduced primary coolant system
inventory and the methods used to assure plant personnel
were aware of the plant .condition (paragraph 3d(3) and Sf).
Manage~ent's involvement during activities that have the *
high potential to affect safety (paragraphs 3d(3), 9d, lOb, * -
10c(2), and 11.
The prompt response to resolve defective anti~rotational
keys in motor operated valv~s (paragraph 6a).
b.
Weakne.sses noted:
1)
2).
3)
The l 1censee found the diesel generator hand switch for the
fuel oil filter in an unspecified position (paragraph 3b).*
The licensee's difficulty in reducing the negative trend
- pertaining to operator errtirs (paragraph~* 3e and 8).
A control room operator appeared to be .overextended while
performing control room activities (paragraph lOc(l)(c)).
c.
The violation for a failure to follow procedures (paragraph 8) ..
d.
The non-cited .violation pertaining to diesel generator
in~perability (paragraph 7d).
e.
The unresolved item pertaining to op~rator errors and the request
to provide a docketed evaluation (paragraph 3e). *
f.
The inspector discussed the potential for providing incorrect
information tb the NRC and potential consequences if an additional
- .example is identified (paragraph 2d).
2
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i.
j.
The two unusual events (two diesel generators inoperable and
excess primary coolant leakage} -were discussed (paragraphs 3b and
3c}.
The licensee response to industry events that could affect the
plant (paragraph 6).
-
Four recent r~actor trips were the re~ults of activities to
improv~ plant performance.
The need_ to proceed cautiou~ly ~ith
future- improvement modifications, and the potential consequen~es
if additional trips were to occur, were dis~ussed (paragraphs 7a
and 7c}.
The need to provide a brief narrative when using "N/A" to d~lete
performance of a test (p-aragraph lOc(l)_(b. k; The auxili~ry building load distribut1on system design and c~lculation were reviewed. The inspector noted that problems with design calculations, such as occurred on the steam generator replace~ent project, did not recur (paragraph 9b}. 2~_ Actions on Previously Identified Items (92701, 92702} a. (Closed} Open Item 255/90018-02(DRP}: SpeDt Fuel Pool (SFP} Leakage The SFP bulk head gate developed a significant water leak in 1990 due to inad~quate nitrogen pressure to the bulk h~ad gate inflatable seals. This- resulted iri a SFP level reduction of approximately -6 feet when the water leaked to the tra~sfer canal. The licensee's evaluation (D-PAL-90:193} documented that the leakage was caused by a faulty nitr6gen pressure regulator. The licensee modified the SFP gate and the pressure regulating system. - The modification included in~tallation of a redundant nitroge~ regulator and a bulk head gate with dual inflatable seals. There have been no additional leaks from the SFP bulk head gate. - b. (Closed) Open Item 255/90021-0l(DRP} as updated in Inspection._ Report 255/9003l(DRP): Safety Injection and Refueling Water Storage (SIRW) Tank l~akage During the 1990 r~fueling outage a leak developed at several of - the SIRW tank bottom plate "pipe to penetrations" welds. The - - licens~e analyzed several failure mechanisms and concluded that the stresses during draining and filling cycles contributed to several cracked welds and the leak. The licensee repaired the cracked welds and implemented a surveillance program to confirm the continuing integrity of the welds. No additional leakage problems have been identified.
3
...
c. (Closed) Violation 255/91018-0l(DRP): Safety related suppott equipment exceeded Limiting Condition for Operation (LCO). PS-0918, "Component Cooling Water Pum*p Discharge Pressure Interlock. Switch," was removed from servite without consideration of Technical Specification applicability. PS-0918 was isolated as part of ground fault trouble shooting activities. The ground fault cleared when PS-0918 was isolated. * PS-0918 iemained isolated for several weeks while the ground fault was iesolved. The safety significance Was minor but indicated that the proper reviews were not performed after the equipment was removed from service .. The failure to perform the reviews resulted in the violation. The licensee's response dated January 19, 1991; to the Notice of Violation acknowledged that timely reviews were not performed after PS-0918 was removed from service. The corrective actions pertained to training of Corrective Action Review.Board (CARB) members regarding_ timely operability determinations. The . inspector routinely reviewed the o~erability determinations of the CARB and has not found additi-0nal examples of this violation. d. (Closed) Unresolved Item 255/92027-l(DRP): Incorr~ct Information Provided to the NRC "The licensee provided incorrect information to the NRC on D~cember 12, 1991,. perta~ning to proposed completion* date~ of ~odificat~ons to safety related circuit breakers. This unresolved item required a docketed evaluation of the causes and corrective actions. The lic~nsee's response dated February 16, 1993,
documented that miscommunication between* the licensing and engineering departments .resulted in the submittal of incorrect. information to the NRC. -The inspector independently developed a time line of the _ -modification and confirmed that miscommunication between-the licensee'~ work groups resulted in the submittal of incotrect infofmation to the NRC. The inspector concluded that the delay in completing the modification did not create a safety hazard . . The licensee's corrective actions included staff training* pertaining to the importance of NRC commitment dates. They also - formalized the definition Of an NRC commitment date. *
Additionally, a guideline for ~reparing submittals to the NRC was prepared.
The inspector eval.uated this item for 10 CFR 50.9, "Completenes~ * and Accuracy of Information," enforcement action.and concluded that the incorrect information was not identified by the NRC or used by the NRC to schedule i nspeet ion activities. This ineets the * criteria specified in 10 CFR 2 Appendix C for not taking 4
enforcement action. The inspector also determined that the licensee did not knowingly provide incorrect information to the NRC. Additionally, on January 1, 1992, enforcement action was issued (Inspection Report 255/91026(DRS)) for several separate examples. of when the litensee provided incorrect information to the NRC: The inspector concluded that .if the incorrect information that .resulted in this unresolved it~m was identif{~d at the time the previous enforcement action was issued, then. this would have been
- an additional example of that enforcement action.
Additionally~ if this item had been included with the enforcement action the severity level would not have changed. ~ased on this, the inspecto~ will not pursue additional enforcement action.
No violations, devf~tions, .unresolved or inspecticin followup items ~ere identified. - * 3. Operational _Safety Verification (71707, 71710, 42700) The facility steady state, shutdown, and startup activities were observed a~ conducted in the plant and from th~ main control room.** Performante of reactor operators and senior r~actor operato~s, shift engineers, and auxiliary* equipment operators was observed and evaluated.
- Included in the review were procedure use and adherence, records and
logs; communications, shift/duty turnover, and the degree of professionalism of control room activities.
. Evaluation, corrective action~ and r~sponse for off normal conditions were examined. This included compliance to any reporting requirements. Observations of the control room monitors, indicators, and recorders wer~ made to verify the operability of emergency systems, radiation monitoring systems, and nuclear reactor protection systems. Reviews -Of surveillance, equipment condition, and tagout logs were conducted .. Proper return to service of selected components was verified. a. General The plant began and e~ded the report pe~iod at esse~tially full* power, although ~ forced outage occurred from April 28 through May 16, 199:3.* b. Unusual Event - Two Diesel Generators Simultaneously Inoperable On April 27, Diesel. Generator (DG) 1-1 was scheduled for preventive maintenance of the air start motors. To facilitate the maintenance activity, DG 1-2 was started and paralleled to the grid. DG 1-2 was loaded to approximately 500 kw when it lost 5
load, wa~ manually tripped~ and d~clared inoperable. DG 1-1 was then vetified. operable by starting it and paralleling_ it to the grid.
The licensee administratively declares a DG inoperable when it is paralleled to the grid because under certairi degraded grid conditions, the DG voltage regulator may cause the DG to tiip and* lockout .. This would require operator action to reset the lockout condition. *When DG 1-1 was paralleled.to the grid, it was
administratively declared inoperable. Since both DGs were simultaneous inoperable (DG 1-1 paralleled to the grid and DG 1-2 unable to maintain load), the licensee declared an unusual e~ent and made the notifications required by the emergency plan and 10 CFR 50.72. The licensee subsequently determined that riotifications to local 6fficials we~e not made * within 15 minutes because qtiestions from the first official * resulted in a communication delay to subsequent officials. The delay was the subject of an internal corrective action document. . . . The reason that DG 1-2 lost load was because of a defective fuel oil pump. The pump had developed a leak at the shaft seal which permitted air to mix wit~ the fu~l oil~ The fuel oil pump was repaired and DG 1-2 returned to service. During ttouble shooting activities ~f DG *1-z, the sel~ctor hand switch for the fue 1 oi 1 filters was found i h an abnorma 1 posit i o_n. With the fuel oil filter removed fro~ servite, the licensee verified that the position of the switch did not reduce the flow of fuel- oil tri the fuel pump.* The inspector agreed with the licensee's contention that the position of the selector hand switch did not affect DG operabilitj. The inspector pointed out to the operations manager and again at the exit interview that since the selector switch wa~ not in the preferred position, it was only fortuitous that the as:-found ,position .of the fuel oil selector switch did not affect operability.* c. Unusual Event - Plant Removed From Service Due to Excess Water* Leakage Frbm Control Rod CCR} Seal Packages . . .
- .
On April 28, the licensee was scheduled to remove. the unit from service because leakoff and temperature measurements for two control rod primary cool ant system pressure boundary seals were* approaching preset administrative limits. The.licensee had be~n monitoring the increased leakoff and temperature measurements .
- since Nove~ber 1992.
Several unsuccessful attempts were made to curb the increasing trend. These included a change in the leakoff sampling.frequency and a change in the exercising frequency. About one-half hour before the commencement of the planned shutdown, the unidentified primary coolant system leak rate was found to exceed the Technital Specification limit of 1;0 gpm by 6
d.
. about 0.15 gpm. The litensee perfor~ed a second leakrate calculation, declared an unusual event, and started the shutdown at 5:37 p.m. The unit was removed from iervice at 10 p.m. The licensee secured fr6m the unusual event on April 3~ at 2:35 a.m. after the plant entered cold shutdown.
- The licensee replaced six control rod drive seal packages, a
primary coolant pump seal cartridge, and performed various other. maintenance and surveillance activities.* The unit was returned to service on May 16, 1993, at 3:02 p.m. * During the post outage heatup~ the inspector verified that the control rod drive seal packages and the primary coolant pump seal cartridge parameters were normal. Shutdown Activities During the outage discussed in the previous paragraph, the inspector made routine tours of the control room. During these tours, the inspector observed that manning requirements were always met, tbe-operators were cognizant of changing plant conditions, the Limiting Condition for Operation (LCO) status board wa~ maintained up-to-date, and the operators were performing assigned tasks iri atcordance with plant procedures. Several of
- the activities observed were:
(1) Plant shutdown to hot standby/shutdown and plant cooldown from hot standby/shutdown per General Operating Procedures (GOP) 8 and 9. (2) (3) Shutdown cooling activities per System Operating Procedure (SO~) 3. Draining the primary coolant system per SOP 1. This activity was performed in two.stages. The first level reduction ~ermitted change-out of the *control rod seal packages. The second stage required draining to reduced inventory for change-out of a primary coolant pump seal_ cartridge..
Prior to commencement of .the second inventory reduction, the inspector performed a review of SOP l*and the licensee's shutdown risk assessment using NRC Temporary Instruction TI 2515/113, "Reliable Decay Heat Removal During Shutdown," as a reference. The inspector verified that both shutdown cooling trains (including mechanical and electrical support sy~tems) were operable, outage activities that could affect shutdown cooling were reviewed and scheduled to minimize the risk to shutdown cooling, and containment integrity was maintained while in reduced inventory . .7.
The inspector ~ttended several ihift briefings pertaining to reduced inventory -activities. These briefings were*. conducted by the shift supervisor, were very detailed, and were attended by senior site managers. The shift supervisor clearly discussed the procedural steps required to obtain shutdown cooling and the lessons learned when shutdown cooling was lost at other sites. The site managers stressed. the safety significance of shutdown cooli~g, emphasized the * importance of self checking, and accentuated the consequences if shutdowp cooling was lost: The licensee ensured that the majority of the plant staff,
- not directly involved with reduced inventory activities, *
were aware of the reduced inventory condition by discussion at the morning meeting. Additionally, reduced inventory was the topic of a news bulletin that received plant wide distribution and was posted throughout the plant. (4) The fr1spector routinely* reviewed the shutdown risk assessment and verified that outage activities did not create any safety concerns. (5) Power escalatioris after synchronization per GOP 5 .. (6) Starting and loading of the Diesel Generator per SOP 22. e. Operator Errors The inspector reviewed the licensee's corrective and preventive actions for several recent operator errors. These error~ were of concern because the number of errors has increased over the past few months. Individually or collectively, none of the errors taused a signifi~ant safety problem. *The general topic - Personnel Errors by Operators - will be addressed as an unresolved item (Unresolved Item 255j93008~0l(DRP)).
As stated. in the cover letter, the licensee was requested to respond to this unresolved item by providing an evaluation of the problems and corrective actibns. Several of the operator errors that have occurred $ince the beginning of 1993 are discussed below..
( 1) (2) On January 6, 1993, both emergency diesel generatbrs were simultaneously and inadvertentlyremoved from service because an operator incorrectly implemented a switching and tagging order. On Jan~~ry 12, 1993, an auxiliary operator misaligned air coolers to the iso-phase bus ducts during his shift rounds . (3) On January 30, 1993, safety injection system surveillance test Q0-21 was commenced without t~o valves in their proper 8
- - i
pre~test alignment. The lineup erro~ was identified, correctedi and the test completed'satisfact6rily. (4) On February 24, 1993, feedwater purity building ai.r compres~or C-903A was incorre~tly removed from service. The error was identified and resolved b~fore the clearari~e was . relea~ed to the mafntenance crew. (5) (6) During the March 10, 1993, biweekly control rod drive testing, an operator tested two control rod drives that were not scheduled to be tested. Performance of the test met the testing frequency specified in the Technical Specification but contributed to an unplanned outage.
On Ma.rch 21, 1993, the spent fuel pool handling machine . malfunctioned due to a personnel error. This error was the subject of the enforcement action discussed in the cdver. letter and discussed in paragraph a of this inspectton report. Jhe licensee has implemented sev~ral.measures to address these problems. They have performed a quality assurance audit of the ri~w sSlf checking arid procedure usage policies, issued an . operations depattment. news letter outlining the above e~amples and requesting that operat6rs review their mistakes and think about
ways to prevent similar errors in the future, and implemented their progressive disciplinary action policy where appropriate . . No violations, deviations, or inspection followup items were iderititied. One unresolved item was identified. 1 t. Maintenance (62703, 42700) Maintenance.activities in the plant were routinely inspected, including both corrective maintenance (repairs) ~nd preventive maintenance.
- MechanicaJl; electrical, and in_stru~ent and control group maintenance*
activities were included as appropriate~ The focus of the .insp~ction was to assure the maintenance activities reviewsd were conducted i~ accordance with approved procedures, regulatory guid~s and industry
- codes or standards, and in conformance with Technical Specifications~
The following items were considered during thi~_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 u~ing approved procedures; and post maintenance testing was performed as applicable; . The following work order (WO) activities were :inspected: a.. WO 24204236 Motor driven auxiliary feedwater pump P-8A coupling inspection 9
b. WO 24300882 Lub~ication of pum~ coupling of motor d~iven auxiliary f eedwater pump P-8A ) c. WO 24301093 Troubleshoot motor and controls of west engineered safeguards sampler pump P-1811 d. WO 24301300 Repair of diesel generator 1-2 fuel oil pump. No viol~tio~s, deviations, unresolved or inspection followup items were identified. 5. Surveillance (61726, 42700) The inspector reviewed Technical Specifications required surveillance testing as _described below, and verified that testing was performed in atcordance with adequate procedures. Additionally, t~st instrumentation wa~ calibrated, Limiting Conditions for Operation were met, removal *and restoration of the. affected components were properly accomplished, and test results conformed with Technical Specifications and procedure requirements .. The results were reviewed by personnel other than the individual directing the test, and deficiencies identified during the testing were properly revi_ewed and resolved by appropriate management . personnel.
The followin~ activities were inspected: a. b. c. d. e. f. MI-1 MI-6 M0-7A-1 Q0-8 Q0-6 Q0-10 NI Power Range, Rod Drop Alarm, Flux Delta T Calibratio~ System Area Monitor Operational Check Emergency Diesel Gen~rator 1-1 ESS Check Valve Operability Test and High Pressure Safetj Injection Flo~ Indicator Verificatioh . Cold Shutdown Valve Test Procedure (Including Containment Isolation Valves) Containment Spray P~mp Operability Test The licensee e~aluated ~urveillance activities Q0-6, 8, and 10 and concluded that performance.of selected sections of these tests. could have a negative effect on shutdown cooling. As a result of this review the licensee delayed performance of these. tests until the end of the forced otitage. At that ~ime the steam generators were available as an alternate ~ourc~ of cooling. No violation~, deviations, unresolved, or inspection f6llowup items were identified.
10
6. Engineering and Technical Support (37700, 92705) The inspector monitored engineering and technical support activities at the site and, on occasion, as provided to the site from the corporate office. The purpose was to assess the adequacy. of these functions in
- *Cdntributing properly to other functions such as operatirins,
maint~nance, t~stirig, training, fire protection, and configuration management. a. * 6efective Anti-rotational *Keys in Motor-Operated Valv~s CMOVl A Region* I II Techni ca 1 Issue Summary documented the discovery of broken anti-rotational keys in some Velan MOVs due to defective material. The inspector met with the MOV system ~ngineer and his supervisor to determine if this problem had Occurred at Palisades. The licensee was aware of the_ prob.lem and had previously contacted the vendor to confirm that the key material may be defective. Additionally, the licensee.was informed that the anti-rotational key will perform*adequately during static t~sting but may fail. during dynamic testing. Palisades ~as 13 Velan MOVs that have a similar "L" shaped anti- ~rotational key installed in the mou~ting flange. The license~ suspects that all 13 valves have the same key materia1. All of the valves were installed in safety related systems. Sine~ the problem .was identified, the -1frensee has performed the following: (1) * _Work Requests were initiated to inspect and replace t~e keys (2) (3) (4) (5) (6) on the 13 MOVs. * The licensee contacted Velan to discuss key material changes ~nd the reason of th~ failure.
- A functional equivalent substitLltiort to* authorize the key
replacement was initiated. An evaluation of pl ant performance and shutd_own risk pl ant performance was initiated. The.licensee contacted other utilities to obtain additional information.
- The system engineer performed a plarit-wide search to
. determine if the problem may be applicable to valves that are not addressed by the licensee'*s GL 89-10 program.* The *inspector found that the licensee's prompt response to this issue was a strength: 11
b. Fiberglass Piping The inspector interviewed the ba 1 ance of p 1 ant* system engineering section head to determine if fiberglass piping was installed at Palis~des. This was in response to a broken fiberglass service water pipe event at another plant. The information was provided to the Nuclear Reactor Regulation License Project Mariager. Fiberglass piping was installed in the cooling tower distribution system that runs along the top of each tower. -The piping is valved to evenly distribute water to each cell of each tower. Pipe sizes range from 24 inches to 60 inches. Sodium hypochlorit~ ~torage tanks T-18A and ~-18Bi as well as Circulating water treatment tank T-44, are all fiberglass* reinforced storage tanks. The piping from these tanks is non- fiberglass. c. NUREG/CR-5822. "Analysis of Thermal Mixing and Boron D~lution in Pressurized Water Reactor {PWR)" The inspector resporded to a Region III request for. information regarding the potential for inadvertent reactivity insertion from a reactor coolant pump restart. NUREG/CR-5822 identified ~ scenario which could result in a reactivity insertion. The inspector was asked to determine if the licensee con~idered this potential transient arid has imple~ented measures to prevent it . . ~ The inspector found that the licensee has.proceduies in place that should prevent such a tran~ient. Standard operating procedure SOP 2A, "Chemical and Volume Control System Charging and Letdown: Concentrated Boric Acid," required that if an operating primary coolant pump tripsi then any dilution in progress b~ immediately stopped. *The procedure also required that at least a primary. coolant pump or a shutdown cooling pump be in operation during a change in boron-concentration.
- *
The transient was also addressed in the emergency operating procedures. The Reactor Trip Recovery Procedure stated that if less than four primary coolant pumps are operating, then*commence emergency boration and establish a 3.75 percent shutdown margin. No violations, deviations, unresolved, or inspection followup items were identified. - 1. Reportable Events (92700, 92720) The inspector reviewed the fcillowing Licensee Event Reports (LERs) by means of direct observation, discussions with* licensee personnel, and review of records. The review addressed compliance to reporting requirements and, as applicable; that immediate corrective action ~nd appropriate action to prevent recurrence had been ~ccomplished. 12 . . ~ .
The five reactor trips that were documented in LERs 255/92034; 92035, 92037, 92038, and 92039 are discussed in the three subsequent
paragraphs. Four trips were the result of modifitatioris to improve plant performance and the fifth was the ~esult of plant aging.- The number of ~eactor trips contributed to a decline in plant performance as indicated on NRC's Quarterly Performance Indicators. This decline was. the subject of two management meetings. Th~ first was held on October 30, 1992, iri the NRC's Region III office. The second was held in the NRC~s Washin~ton DC office on November 30, 1992. a. (Closed) LERs 255/92034, 255/92035, and 255/92039: Reactor Trip caused by Loss-Of..:Load Resulting From Unstable Voltage to the * Turbine Control Computers. Three full power t~tbine/teactor trips otcurred on July 1, July 24, and October 30, 1992, when in~plant voltage transients mo~entarily interrupted power to the tutbine control computers. The turbine control computers were designed to prevent a turbine over~speed event. Wheneve~ the computers are unable.to monitor turbine speed, the turbine is tripped automatically. The reactor trips_ and subseque~t plant r~sponses were discussed in Inspe~tion Reports 255/92018(DRP), 255/92022(DRP), and 255/92023(DRP). The turbin~ control co~puters were added du~ing the ~revious refueling outage to erihanc~ turbine reliability. The modification provided redundant computers, redundant power supplies, and the capability to switch from a primary to an alternate- computer or power supply. Additionally, the licensee elected to keep the turbine control computer power supplies independent of safety related power supplies. This assured that safety and non-safety related power supplies were not cross connected~
However, .this modification resulted in unfiltered.and unstable - powe~ to computers that were designed to functi-on with very sta~le
- power supplies.
The modification package documented that problems with similar turbine control computers installed at other plants* were evaluated and resolved before the modifications were implemented. However, the installation ~ackage did not document that differences in the installed configuration were evaluated. (1) The July 1, 1992, -turbine/reactor trip The licensee did not tecognize the significance of a switchyard electrical transient that preceded the turbine/reactor trip. The transient was caused by* a ground fault that cleared several cycles before the turbine trip. The licensee concluded that the transient had cleared but incorrectly concluded that the transient did not cause the computer fault. The licensee concluded that connecting * cables in the control computers data processing units were not properly connected, vibrated loose, and fa~lted the computers. This tripped the turbine and generated a loss-of-load signal to the reactor protective system. The 13
licensee was able to duplicate the computer fault by "wi.ggl ing"- the connections. The control computers were located in a high vibration area, however, as shown in the . subseq~ent paragraphs, the most probable cause was *the voltage transient. The licensee acknowledged and documented this conclusion in a subsequent LER. The licensee did not correctly identify the root cause but . may have identified and may have resolved a future turbine trip mechanism - the loose connectors. (2) The July 24. 1992, tur~ine/reactor trip The licensee concluded .that this trip was caused by an in-plant voltage transient generated during the performance of an unrelated surveillance test. The transient caused a momentary loss of the turbine control computers. The licensee upgraded the computer power supply switching cards and added a single uninterruptable power supply (UPS) to the power feeds for selected turbine control computers~ The UPS provided a filtered power supply that could control rapid ~6ltage transients and provided'a limited.battery supply to. ~ermit operator response if power was lost to the tontrol computers. The licensee's post modification testing
confirmed that the UPS resolved rapid transients and helped manage a tot a 1 loss of power to the contro 1 computers.
The licensee did not consider or evaluate the consequences
- of a degraded or slowly degrading power supply to the
control computers. Failure to test for degraded voltage contributed to the October 30, 1992L reactor trip. (3) The October 30, 1_992. turbine/reactor trip
- The licensee determined that the supply voltage transformer
to the DEH computers was consequently degr~ded by approximately 10 percent because the taps had been improperly set. This caused the UPS internal logic to remove_the UPS from service and permit degraded and . unfiltered-~ower to the control.computefs. The d~graded and
- unfiltered power supply eventually faulted the computers and
caused the trip.
This was resolved by changing the transformer tap setfing* which increased the nominal supply voltage. Additionally the licensee revamped the power scheme to the control .. computers by replacing the internal computer power supplies, increasing the battery size to the UPS, installing voltage regulating transformers to the control compute~s not powered by the UPS, and rerouting the power feeds to provide redundant power supplies to the control computers. 14
r,. b.
(4) _ The inspector con'cluded that: (a) The licensee had evaluated and resolved problems encountered at other facilities prior to installation of the turbine control computers. However, the licensee did not evaluate the-differences between.the (b) (c) ( d) - (e) . (f) (Closed) install~d Palis~des con~iguration and the.inst~lled configuration at other power plants.
The licensee did not evaluate the power scheme prior to i~stalling the turbine control computers or prior to installation of the UPS.
The current power ~cheme was redundant from load ceriter 14. The licensee recognized that the power supplies to load center 14 were not redundant. The licensee was evaluating modifications to enhance the* power scheme. The oper~tor and plant response to the reactor trips were uncomplicated which speaks highly of the material condition of the plant and the training provided to. the operators.
- *
The current power scheme was tested several times when unplanne<;I ex_ternal line faults caused internal plant transients. In all cases the turbine control computers remained in service. The licensee reorganized the Design Engineering Department following the 1990 steam generator replacement outage and relocated the department to the site~ The inspect6r was*unable to determine if the reorganized dep~rtment iriherited the DEH modification or was the sole sponsor of the modification. LER 255/92037: -Reactor Trip_Caused by Low Steam Generator Water Level Resulting from a Broken Air Line on a Main Feedwat~r.Regul~ting *Valve. ' . . . . A full pow~r plant trip. occurred on August 14, 1992, wh~n the feedwater regulating valve to the "A" steam generator malfunctioned. The malfunction was the result of a broken air line to the ~alye actuator. Thi~ permitted the valve to drift partially shut, resulted in ~ reduction in feedwater flow, -reduced w~ter level in the "A" ste~m gener~tor,* and a reactor trip. The * air line was replaced and the air lines to other valves located in the turbine building were inspected before returning the unit to service. The reactor trip and subsequent plant response were ~iscussed -in Irispection Report 255/92022(DRP). , 15
c.
(Closed) LER 255/92038: Reactor Trip Caused by a Loss of the Preferred Bus Y-20 Coincident with a Blown Fuse in a Second Channel of the Reactor Protective System.
- The full power plant trip on August 25, 1992, resulted from a
. ~al function of preferred ac bus Y-20, *co{ncident with a blown fuse . in another portio~ of the reactor prot~ctive syst~m (RPS). The _reactor trip and subsequent plant response were discussed in Inspection Report 255/92022(DRP). The malfunction was caused by an inc~rrectly wired transformer in Y~20. The wiring error *occurred during. corrective maintenance performed in 1986 and resulted in accelerated aging of v~20. The :RPS consists of six logic ladders (AB, AC, AD, BC, BD, and CD) representing the 2*out of 4 logic combinations. Each logic ladder has two auctioneered power supplies which are powered from
separate preferred ac power sources .. One side of the BC logic matrix ladder is powered by the Y-20 bus while the other side i"s powered by the Y-30 bus. A fuse blew in the BC matrix power* supply which is powered from the Y-30 inverter. The exact time.* that the fuse blew is unknow~. Since the power supplies are auctioneered, either matrix power supply can be in service .. When a matrix power .supply is out of service it is not ann~riciated. The fuse was determined to be undersized and could have blown when the power supply was first placed into service following the refueling outage, after some duty time, or when Y-20 malfunctioned. The* power supplies (12 total) were replaced du.ring the previous refueling outage as part of a material condition improvement: The. licensee's post trip evaluation determined that the undersized fu~e was installed at the vendor's manufacturing facility. The fuses in. the other power supplies were inspected and found to be properly_ sized.
The site inspection staff and a Region IJI inspector evaluated this LER and concluded th~ following . . 1) The licensee's reteipt inspection requirements for the logic 2) . ladder power supplies were evaluated. * The i nsp.ector concluded that the pre- and post- installation activities in combination with the vendor's quality inspection program were adequate to confirm operability of the power supplies. The inspector concluded that disassembly of the power supplies during receipt inspection was not re~uired; The current maintenance programs are significantly different from the program that was* in place when the Y-20 transformer wiring problems occurred approximately 8 years ago. The difference included the. use of system engineers, use of w6rk 16
d. . . order planners, and implementation of a computeri~ed work order system. (Closed) LER 255/92-036: Inadvertent Isolation of Fuel Oil to One Cylinder of the 1~1 Emergency Diesel Generator. The subject of this LER was reviewed in Inspectiori Reports 255/92022(DRP) and 255/92023(DRP) . . (1) General On Augus~ 2, 1992, the l~censee was performing M0-7A-l~ "fmergency Diesel Generator (DG) 1-1 (K-6A)," when they observed that cylinder BR indicated a low exhaust t~mperature of about l70 degrees F. This was well below the expected temperature of approximately BBQ degrees F .. The licensee decl~red the DG inoperable and dispatched an operator to.check the fuel rack to cylitider BR. An auxiliary operator found the fuel rack.to cylinder BR was "latched out" or "shut off." The auxiliary operator opened the latch and the surveillance test was completed satisfactorily. The fuel racks were* locked out 33 days earlier.to perform compression testing per ~0-7A-l. FollowiDg.the compression test, the licensee ran the DG and documented that the exhaust temperatur~s were ~orrect. AdditionallY, the DG was tested at least twice following two subsequent reactor
tr"ips. Although this.-testing did not require verification * of cylinder exhaust temperatures, the licensee's records did not doc~ment anY operational problems. (2) Potential Duration of Inoperability (3) The lfcensee back dated the ~noperability a total of 33 * .<;Jays .. This was the elapsed time* since the compression te~tirig was satisfactorily performed o~ June 30, 1992. This meant that the inoperability condition could have existed longer than the allowable outage time of Te~hnical Specification 3.7:2.i. Technical Specification 3.7.2.i permitted an .out-of-service time of 7 days, o~ required a plant shutdown. Root Cause and Corrective Action The most likely cause for this event was a failure to implement post compression restoration steps of M0-7A-l .. Normally the individual fuel rack latches are stored in the 116 o'clock" position. The latch for this fuel rack may have stuck in an "intermediate or 12 o'clock" position. The .. licensee tonfirmed that it is possible for fuel rack latches to stick in an intermediate position .. Vibrations may have 17
- ~~used the latch to vibrate from an intermediate position
- into the "lockout or 3 o'clock" position, thereby isolating
the fuel to cylinder SR. This apparently occurred* at some* time during the four times that the DG was operated since the compression test.
The li~ensee's corrective action~ include~ revising*M0-7A~l and M0-7A-2, "Emergency Diesel Generator 1-2 (K-6B)." The revisions included specific directions for proper fuel control rack latch disengagement and caution statements explaining consequences of not completing this propeily. Operator training ~as alio held. (4) Previous Fuel Rack Operability Problems (5) While evaluating the licensee's past performance to address the.criteria of Section VII.B of the enforcement policy (10 CFR*2 Apperidix C), th~ inspector revisited anot~er cylinder exhaust temperature problem which was documented in NRC Inspection Report 255/90039 (DRP). * On December 18, 1990, maintenance mechanics were performing . preventive maintenance on DG 1-2 when a mechanic unknowingly place~ the latch to cylinder 2L in the locko~t position. He thought he was placirig it in the correct position. The licensee found this condition during post-maintenance testing and*prior to performing the operability run. Operators placed the latch in the correct position and later successfully completed the operabi-li ty run.
- The *i~spector determined that the corrective actions for the
- December 18, 1990, event were adequate.
They included
- training of the_maintenance methanics and posting of warning
Signs in both DG rooms. The inspector found the corrective actions for the 1990 event could not reasonably have prevented the 1992 event. To date, there have been no other
- instances of mechanics incorr.ectly positioning fuel rack
linkages,
Potential Enforcement Action This event was evaluated for escalated enforcement by Region III and the Office of Enforcement. Their evaluation determined that escalated enforcement would noi be pursued because the post maintenance test performed subsequent to the compression test did verify operability of* the fuel racks. Because of that test, the length of time that the DG was inoperable prior to discovery of the rue l. rack .
mispositioning was indeterminate. Therefore, it was not possible to establish that there was any_ viol~tion of . Technical Specification 3.7.2.i. 18
A violation for not implementing the restoration steps of M0-7A-l was not cited since the criteria specified in
- Section VII.B.2 of the "General Statement of Policy and
Procedures for NRC Enforcement Actions~** (Enforcement Policy, 10 CFR Part 2, Appendix C),.were satisfied. e. (Closed) LER 255/92017: Fuel Assembly Partially R~moved From ihe Core During Upper Guide Structure Lift.* ' . On February 29, 1992, while removing the upper guide structure from the vessel, a fuel assembly at location Z-11 remained attached to the bottom of the upper guide structure. The licensee declared an unusual event, stopped the lift, and implemented . contingency plan~ to secure, free, and lower the fuel assembly back into the core. The licensee subsequently determined that the upper guide str~cture alignment pins for location Z-11 were bent . and caused the fuel assembly to stick during lifting.* The pins were straightened before the upper guide structure was installed during vessel restoration actfvities.
NRC inspections were *documented in Inspection Reports * 255/92006(DRP) and 255/92015(DRP). In addition, this event wa~ the subject of enforcement action documented in Inspection Report 25S/92015(DRP). . . . . . f. (Closed) LER 255/91020: Inadequate Documentation - Environmental. -Protection P.l an. On November 12, 1991, the NRC identified to the licensee that they had not properly documented Unreviewed Environm~ntal Question (UEQ) determinations for plant modifications. This was the subject of enforcement actio~ documented in Ihsp~cti-0n Report 255/91024(DRP). The licensee subsequently determined that UEQ determinations had not been properly documented since February 1987. In February 1987, Administrative Procedure 4.22, "Nonradiological * Environmental Program," was revised and deleted the for~ which provided documentation of an UEQ evaluation in compliance with the Environmental Protection Plan (EPP)._ The plant had relied on the Corporate Environm~ntal Department .to determine if a test, change, procedure or modification could involve an UEQ. The li~ensee stated that the Corporate Environmental Department was responsible for compliance with federal, state, and local environmental regulations: However, they w~re not trained or knowledgeable in nuclear regulatory
- compliance.
The cause of this event was a procedure inadequacy. The licensee's corrective actions were: (1) suspend processing of environmental impact reviews u~til Administrative Procedure 4.22 19
..
was revised to include steps to ensure the environmental requirements of the EPP are met for determining the involve~ent_of a UEQ; and (2) review all evaluations of the involvement of UEQi * performed since February 1987 and perform required evaluations as necessary. The inspector verified that Administrative Procedure 4.22 was properly revised. * He also reviewed several design changes and confirmed that required evaluations were performed. * * g. (Closed} LER 255/92004 (original and supplement dated March 6, 1992, and ~ay 15, 1992): Potential Loss ~f Containment Integrity Due to the Failure of the Emergenc*y Escape Airlock Equalizing Valve. The LER w~s superseded by LER 255/93003. * The licensee concluded that the equalizing valve stuck partially open becaOse the valve stem lubricant partially dried and became tacky. The licensee cleaned the equalizing valve during the 1992 refueling outage and lubricated the stem with the vendor recommended lubric~nt. The equalizing valve was returned to service and operated properly during.post maintenance testing . . The equalizing valve failed a second time during airlock testing perfrirmed on March 6, 1993. The second failure was discussed on LER 255/93003. That LER remains open to be evaluated when the valve is disassembled during the upcoming 1993 refueling outage. One non-cited violation was identified. No deviations, u~resolved, or inspection followup items were. identified. 8. Spent Fuel Pool Handling Machine CSFHM) Malfunction Due to a Personnel Error a. General During preplanned moves of spent fuel (to s.upport the up-comi.ng 1993 refueling outage}, the SFHM malfunctioned due to a personnel error .. The malfunction occurred after an operator had inappropriately ~sed the SFHM "Override Key Switch" in an unsuccessful attempt to seat and ungrapple a fuel assembly in a* storage location. When the operator attempted to remove the fuel ~ssembly from the storage location, several inches of the ~ain * cable unwrapped from the SFHM drum and wrapped arou.nd the motor shaft. The ~able remained attached to the drum but allowed the fuel asiembly to drop approximately six inches. The fuel ass.embly was suspended approximately six inches above the bottom of the spent fuel pool _(SFP}; This occuried at approximately 2:30 a.m. on March 21, 1993. The licensee verified that the fuel assembly was not damaged and confirmed that there was no release of radioactivity. 20 '
- .
b. The shift supervis~r directed that power to the SFHM be secured and he suspended activities in the SF~. The shift engineer verified that the emergency plan had not been entered and that a - 10 CFR 50.72 telephone notification was not_required. A system engineer, a system engineering section ch]ef, the shift - ~upervisor~ and a field representative for the refueling machine examined the config~ration and concluded th~t the fuel assembly was in a safe configuration.
NRC activities The senior resident inspector was notified at approximately . 4:30 a.m. and responded to the site. He attended the licensee's technical briefings, interviewed the operator who performed the fuel rnoves, reviewed the procedure controlling the activity, and independently verified that _the fuel assembly was in a safe configuration. Based on interviews with the operator performing. the activity, the inspector concluded that personnel error was* a primary cause. The inspector informed the operations superintendent at approximately 9:30 a.m. that perspnnel error was a primary cause of thi~ event. c. * Spent Fuel Pool Handling Machine Opetatinq Information d. The SFHM can be operated in semi-automatic mode, in manual mode with the computer interlcicks functioning~ or in manual mode with the computer interlocks and monitoring capabilities bypassed! The manual mode (with the computer in.service) permitted _manual operation of the bridge, trolley and hoist within operating zones
- that were preprogrammed into the computer.
The hoist is-only operated in manual and is normally operated with the co~puter _ controlling the speed of the hoist, monitoring.the load, the** position of the fuel assembly, the upper grapple operating zone (UGOZ) and lower grapple operating zone _(LGOZ). The se~i-automatic mode permitted computer ~ontrol of the direttion, speed, and oper~ting zones of the bridge and trolley. . . . . ' . The system (bridge, trolley, and hoi~t) can be operated without the com~uter b~ use of the "Override Key Switch~" The "Override Key Switch" bypassed all of the limit switches and operating zones. *This mode of operation* provided the operator with a means to safely store a fuel assembly in the event of a computer malfunction. The operating procedures for the SFHM were contained in Section 7.6 and Attachment 5 of System Operating Procedure (SOP) 28, "Fuel Handling System." Event and Violation While lowering fuel assembly XF-4 into storage location QW-35, the operator observed normal weight of 1300 to 1400 pounds. When the 21
fuel assembly was approximately one and one half inches from seating~ the underload interlock activated and ~utomatically stopped the hoist. In this c~se, th~ unde~load conditibn identified that the fuel assembly alignment pins were resting on the bottom of the storage location ,but not entering the alignment. holes of location QW-35. Two alignment pins (approximately one and one half inches in length) are located on the bottom of each fuel assembly. Each storage location has four alignment holes. A fuel assembly was considered properly sea_ted if it traveled the programmed distance and the alignment pins entered the alignment holes. When,these conditions w~re established, the lower grapple operating zone (LGOZ) position switch would activate, the underload condition would activ.ate the cable slack limit switch, and permit a manual ungrappling of the fuel assembly. When the _underload condition was r.eceived, the oper~tor raised the fuel assembly several inches, lowered the fuel assembly, and attempted to reseat the fuel assembly. The underload condition was received a second time. At that time, ~he operator activated the "Override Key Switch" and unsuccessfully attempted to lower the fuel assembly to the_LGOZ and ~eat the bundle. The operator then removed the fuel assembly from storage location QW-35, .. realigned the fuel assembly, and attempted to reseat the fuel assembly. The operator was unable to se~t the fuel assembly and again attempted to seat the fuel assembly by use of the "Overr~de Key Switch." The operiior-used the "Override Key Switch without permission of the shift ~uperviso~ which was a vi6lation of step 1 to Attachment 5, "K~y Operated Override Switch Operating Guidelines" of SOP 28. Step 1 stated, "Use of key operated override switch shall be at the direction of an SRO." Technical Specificatiori 6.8.1.b required implementation of ~rocedures for fuel handling activities. SOP 28 implemented this requirement .. Failure to cofuply with step 1 of Attachment 5 to *sop 28 is a * ~iolation of Technical Specification 6.8.1.b (Violation 255/93008- . 02 (DRP)). The operator removed the fuel assembly from the storage location,
- rotated the fuel assembly 90 degrees, and successfully seated the
bundle in storage location QW-35. The operator was not able to* ungrapple the fuel assembly. The operator then attempted to remove the fuel assembly with the intention of rotating the fuel assembly 90 degrees. When the operat6r raised the fuel assembly approxim~tely 12 inches, ~everal inches of the main cable unwrapped from the drum and wrapped several times around the drive shaft. This permitted the fuel assembly to drop approximately *six in~hes and become suspend~d approximately six inches above the bottom of the SFP.
22
e. Review of SOP 28 The inspettor reviewed SOP 28 to determine if the p~ocedure was adequate for the activity performed. The inspectoi found that th~ procedure was easy to follow and addressed the normal or routine activities. The inspector questioned i.f the procedure ade~u~tely addre~sed off normal conditions. For example, th~ procedure was .silent with respect to the underload condition.
Th~ procedure required shift supervisor approval prior to use of the "Override Key Switch." It also provided -specific criteria pertaining to its use. The violation was based on the operator's failure to obtain shift supervisor permisiion prio~ to use of the Override Key Switch. The inspector also*nrited that if the . opeiator had reviewed the procedure he would have not proceeded beciuse use of the override key switch for this purpose was not addressed. f. Recovery Activities The system ~ngineer and the SFHM vendor implemented a course of action that included installation of a clampin~ device on t~e main cable, attaching the clamping device (via a chain fall) td the
overhead crane, and suspending the cable and fuel assembly_ from the overhead crane. The cable and drum were manually detensioned * and the cab1e rewound on th~ drum. The drum was manually
tensioned,. the clamping device removed, the bundle lowered arid the grapple disconnected from t~e fuel assembly ~ith the aid of a long_ . reach pole. One violation and no deviations, ~nresolved, or inspection followup items were identified. g. Drj Cask Storage Operations (42700, 86700, 37702, 37703, 71707) a. Decontamination of Multi-assembly Sealed Basket (MSB) - . . . . . . The inspeetor observed part of the decontamination of the MSB and the MSB Transfer Cask (MTC) according to procedure T-FC-864-01, "Preoperational Test Procedure for Loading and Pl~cing the
Ventilated Storage Cask into the Storage," The inspector found
- that the evolution was properly coritrolled.
Information provided during the pre-job briefing included precautions, a review of the procedural steps, and identification and responsibilities of key management, technical, and operations personnel. The briefing was effecti~e and met the objectives of. the procedure. The licensee. performed this.activity to verify that the gap between the MTC and the MSB could be decontaminated to a level* less than 2000 cpm/lOOcm2 smearable. The decontamination was * 23
accomplished by fl ushirig the. gap with pure water as it was 1 ifted out of the spent fuel pool; Prior to the flushi the licensee measured the contamination level and found the_ highest contamination.level to be 800 cpm/lOOcm2 , well below the. acceptance criteria. The gap was then fl~shed_to further reduce the contamination levels. . . . . . The decontami na.t ion of the MSB/MTC gap was the second ti me this phase of the preoperational procedure was performed. The licensee had a more difficult time decontaminating the MTC/MSB a fe~ weeks .earlie~. They suspected this was partly due to the e~tended amo~nt of time (approximately 4 weeks) the MTC/MSB r~mained in th~ * spent fuel pool during reforbishment of its transport system. They also suspected a problem with the shim material used to maintain the gap. _The shims provide radial alignment of the MSB withirr the MTC. The licensee changed the shim material from * carbon.steel to stainle~s steel. The carbon steel material made .decontamination more difficult because corrosion products formed due to reaction with the borated water in the sp~nt fuel pool~* . . For this current test, the MTC/MSB was in the spent fuel ~ool about 24 hours prior to its decontamination. During actual fuel loading, the MTC/MSB is expected to be in th~ spent fuel pool less than 24 hours. b. Review of Load Distribution System CLOS) Th~ LDS was installed in the auxiliary buildin~ track alley to * . evenly distribute the load of the ventilated concrete cask t6 the floor and rooms located below the track alley. Portions of the. following calculations were reviewed for compli~nce with NRC requirements and conformance to licensee commitments: ( 1) (2) (3) . EA~F-C-864~01-9, "Structural Analysis* of the Palisades Track*. Alley Bridge for VSC-24 Lbading Operati-0n,ri Reviiion 0, April 22, 1993. EA-FC-864-020, "Weight Calculation for VCC, MTC; and M~B, 11 Revision l~ February 3, 1993. EA-FC-864-011~ "Evaluation of MSB for Drop Loads for a Hypothetical Drop on the LOS in Track Alley," Revision 2, April 9, 1993. The analyses were well organized with all assumptions adequately identified. The evaluation of the LOS used a simplified analytical approach and a~plied bounding l~ads to the structure at worst-cas~ locations. Point loads wete us~d instead of*
distributed loads, which added to the ~onservatis~ of the * calculation. No technical deficiencies were noted during the
- review of the calculation.
24
.. '~ . l "**
.. All calculations provided by the transfer cask vendor were reviewed by-either cognizant licensee personnel or by a third
- party consultant.
The design verification documentation associated with these reviews indicated that they were technically thorough and comments were adequately resolved. . I ri some cases,
.alternate calculations we~e provided to demonstrate the adequacy of the initial calculation and to preclude any q~estions in that regard. The NRC inspe~tor briefly examined the install~d LDS. Grout installed between the baseplates *and road surface had several.
- cracked areas. These were restricted to the portion of the grout
. that extended beyond the edge ~f the baseplates.and did not appear to be structurally significant. In general the overall
workmanship and quality of welding were good. No significant deviations were noted between the design drawings and a~-built structure.
- *
.*
During previous inspections .that evaluated management of design changes, specifications, and calculations, a number of problems were identified pertaining to control of contractors. The
inspector's limited review of the LDS did not identify any . additional probl.ems in that area. - c. Public Demonstrations The inspector observed two public demonstrations protesting* against the licensee and one supporting the liceniee in the u~e of dry cask storage. These demonstrations were spon~ored by several local groups and held at the plant access road. The .*. dem-onstrat ions were peaceful, did not obstruct traffic, received
- exten~ive media coverage, and averaged approximately 100 persons.
~er demonstr~tion. d. Plant Review Co~mittee*{PRC) The inspector attended the May 1, 1993, special PRC .meeting.* The** purpose of the ~eeting was to summarize the status rif the Dry Fuel Storage (DFS) project. This was accomplished by performing a review of the pre-operational test results, summarizing the status
- of the engineering design packag~s. discussing the Certification
- of Compliance (CofC), reviewing the ~ask *1oading schedule, an4
evaluating outstanding licensing issues. The PRC concluded that there were no unreviewed safety questions. . . The items discussed were presented to plant management by knowledgeable individuals. These individuals were able to discuss the criteria used to deter~ine whether items were successfully completed. In addition, these individuals were able to discuss the lessons learned and the technical merit of the solutions to any problems. *
25
. ' The inspector verified that the -PRC composition ~et Technical Specification manni.ng requirements and that a voting quorum was present..
e. Legal Activities On May 4, 1993, a motion for a Temporary Restraining Order was filed by the Attorney General for the State of Michigan, the Lake Michigan Federation, and several land owners with property on the shore of Lake Michigan.and adjacent to the plant. The motion was for an order temporarily restraining the NRC from permitting outside storage of spent fuel in VSC-24 casks at the Palisades_ Nuclear Power*Plartt. On May 8, 1993, the Judge ruled that his court did not have jurisdicti~n. On MaY 13,.1993, a motio~ for an Emergency Stay was.fil~d ~ri Appe 11 ate Court by the same personnel listed in the preceding paragraph. The Emergency S~ay w~s for an imm~diate stop work order for using the VSC-24 casks at Palisades Nuclear Power Plant. On May 17, 1993, a panel of three judges at the Federal Court of appeals_ in Cincinnati, denied the plaintiffs' request for an Emergency Stay.
No violations, deviations, unresol~ed, or inspection followup items were identified. 10. Loading of the First and Second Dry Storag~ Casks (71707 ~nd 42700)
- Loading of the first cask started shortly after midnight on May 7, l993.
Loading of the second cask started on May 12, 1993. The site NRC inspection staff, NRC inspectors stationed at other plants, Region 111- based inspectors, and NRC* Headquarters-based inspectors implemented round-the-clock cover~ge while the licensee loaded fuel {n the casks and
- * transported the casks to the storage pad. *The activities listed below
were inspected. a. The :inspector performed a limited review of the following procedure using 10 CFR 72; the NRC Safety Analysis Report_ for the Ventilated Storage Cask System - dated April 28, 1993; CofC Number . 1007 - ~ated May 3, 1993; and the American Society of Mechanical Engineers (ASME) Code as references.
- *
. . ( 1) FHS-M~32 11 Loading and Placing the VSC into St_orage 11
The inspector verified that the procedure required at least two vacuum pressure drying times with an acceptance criteria of less than 3 mm Hg vacuum, a helium leak test at 22.1 psia, and a final helium pressurization to 14.5 psia. The inspector also verified that the procedure imposed a minimum* temperature for lifting the MTC and moving the VCC, and that it imposed a handling height limitation for the MTC and VCC. 26
( 2) . (3) During interviews with the site licensing personnel, the inspector was informed that the MSB drain time w~s _admtnistratively reduced.based on the ambient temp~rature of the water in the spent fuel pool. FHS-M-34. "Unloading the Multi-Assembly Sealed Basket" FHS0-17 "Multi-Assembly Sealed Basket Loading Procedure" The inspector interviewed several reactor engineers and the reactor engineering section chief to confirm that the spent fuel assemblies stored in the first two casks met the specification of the CofC. Several questioris were asked pertaining to the heat loading ~f the first cask and integrity of the fuel cladding for the fuel assemblies pl~ced in both casks.* Th~ selected heat loading was
- appro.ximately 12 kW, which was the maximum heat loading
, available* for the required post irradi~tion tim~. * The. integrity of the fuel cladding was confirmed by visu.al inspection. Additionally, the fuel ass~mblies of the first two casks were "sipped" when they were temoved from the core in 1986 to ens~re they were not leaking .. (4) COP-27 "Spent Fuel Pool Chemistry Operating Procedure" The inspector verified that the procedure required independent determinatibns of the spent fuel pool boron concentration.
(5) FC-LID and SM~LID "CPCo Welding Procedure" The inspector reviewed these Weld Procedure Qualification
- Becords (PQRs): FC-LID-A, FC-LID-8,
SM~LID-C, .and SM-LID-D; these certified material test reports: weld wire E71T~l, * heat #32039; weld electrode E-7018~3/32, heat #91232; .and* weld electrode E-7018-1/8", heat #T20658. In addition t_he inspector reviewed these welder certifications: Flux.core automatic welding (FCAWJ: 38, 54; Shielded metal arc welding- (SMAW) 38, 54, 7H, and 7B. The documents reviewed.* complied with the requirements of the AS~E Code, Se~tion III, .NC 4000 and Section IX, 1986 Edition, Summer 1988 Addenda~ b. The inspector attended the prejob and ALARA briefings for the following:
(1) Movement of the MTC/MS8 into the sperit fuel *pool. (2) Movement of the MTC/MSB from the spent fuel pool to the cask
- washdown pit.
27
.... (3). Welding of the shield lid and structural lid to the MSB . (4) Movement of the MT~/MSB from the cask washdown pit to the VCC and placement of the MSB into the VCC. (5) Movement of the VCC to storage. The pre-job and ALARA briefings were very detailed, conducted by the cognizant supervisor, and attended by senior site managers. The cognizant supervisor clearly discussed the precautions, the procedural steps expected to be accomplished, and identified the responsibilities of the key personnel. The site managers stressed the safety significance of performing the activities in a well con.trolled manner, and stressed the importance of both personnel and plant safety. Th~ inspector noted that the managers did not stress the schedule for completing the acti~ities, and in several instances stated that completion of a quality and safe activity was more.important then meeting schedular predictions. c. The inspector observed the following activities: -{ 1) Placement of fuel into both MSBs per FHSO 17. {a) . {b) {c) The inspector verified that operators.performed dual' verification of the identity of each fuel assembly before insertion in the MSB.
The inspe~tor reviewed the official topy of FHS0-17, and noted that the section pertaining to checkout of the refueling machine had been marked "N/A". No explanation for the "N/A" was provided. After some review it was determined that the "N/A" was appropriate because the activity was performed by another procedure. The* irisp.ector noted that a 'brief explanation for the "N/A" appeared appropriate. A brief explanation was subsequently included. While loading the second cask, the inspector noted that the control room operator directing the fuel moves and maintaining the status board was involved in other duties and appeared to be having difficulty ** completing all assigned duties. This was discussed with the shift engineer who provided additional assistance.
(2) Welding per FC-LID and SM-LID. The inspector observed welding of the MSB shield lid for the- first MSB. The inspector verified that all welding essential. variables were performed in accordance with the applicable welding procedure and that nondestructive examination {liquid penetrant) was performed at the required 28
(3) (4). intervals. The welders were qualified for the process employed, welding materials were identified, and.stored in accordance with ASME Code requirements. The licensee assigned a welding engineer to*observe and assure that welding activities were performed to applicabl~ requirements. The inspector c6ncluded that the licensee implemented a we 1 ding program that was cons.ervat i ve and assured that the .MSB welding was accomplished to the required standard~ and codes.
The licensee's management attention was_duly noted during the* inspection. Movement of the MSB/MTC from the wash down pit to the spent fuel pool, from the spent fuel pool to t~e wash down pit and to the VCC, and placement of the MSB into the VCC per FHS-M-32. These activities were observed for each MSB. Movement of both VCCs to the storage pad and placement on the storage pad per FHS-M-32. * d, The inspector concluded the following: (1) That the procedures reviewed fmplemented the applic~ble portions of the certificate of compliance. (2) Management tnvolvement was highly visible. (3) The ease that the licensee placed fuel, welded, and transported the casks to storage demonstrated that the preoperational test progra~was effective and wel.l managed. No vi~lations, deviations~ unresolved, or inspection follo~up items were i dent ifi ed. 11. Quarterly Management Meeting A Quarterl~ Management meeting was held at the Palisad~s- site on May 17~ 1993, with the personhel indicated in paragraph 10 in attendance. Th~ following items were discusse~: A discussion of the current plant status, including dry fuel ~torage operations.
- A review of plant operations from*November 8, 1992, through April
28, 1993 ~ 29
A discussion of the problems, caus~~. and correctiv~ actiohs .* associated with the turbine generate~ digital electro-hydra~lic
- control system ..
- . An ov~rview of the 1993 refueling outag~ scope. . . . A discussion of- N~C inspections as they impact billing of the licensee and on the lic~nse~'s refueling outage. No violations, deviations, unresolved, or inspection followup items were i dent ifi ed.*. 12. Persons Contacted Consumers Power Company
- D. P. Hoffman, Vice President, Nuclear Operations
- ~. B. Slade, Plant General Manager
-
- T. J. Palmi~ano, Plant Operations Manager
. D. J. VandeWalle, Mech/Civil/Structural Engr. Manager
- R. D. Orosz, Nucl~ar Engineering & Constrlicti~n Manager
- D. W.
Roge~s, Safety & Licensing Director
-
- K. M .. Haas; Radiological Services Manager
J, L. Hanson, Operations Superintendent R. B~ Kasper, Maintenance Manager
- .K. _E. Osborne,
Sy~tem Engineering Manager . W. L. Roberts, Senior Licensing Engineer K. A. Toner, Electricaljl&C/Computer Engineering Manager Nuclear Regulatory Commission {NRC)
- W. l. Forney,* Deputy Director, Division of Reattor Projects, Riii
- W. M. Dean, Acting Director, Project Dir~ctorate III-1, NRR
- W. E .. Scott, Performance and Quality Evaluation ~ranc~, NRR
- W. D. Shafer, Chief) Reactot- Projects Branch 2, ~III
- A. H. Hsia, Licensing Project Manager, NRR
- E. R. Schweibinz, Senior Project Engineer, Regfon III *
- J. K. Heller, Senior Resident Inspector.
- #D. Passehl,. Resident ~nspector
- Denotes some of tho~e pr~sent at th~ m~nagement interview on May 24,
1993.
- Denotei some of those present at the quar~erly management.meetirig held
on. May 17, 1993. Other members -of the plant staff, and several members of the contract security force," were also contacted during the inspection period. 30 }}