IR 05000255/1992018
| ML18058A986 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/27/1992 |
| From: | Jorgensen B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18058A987 | List: |
| References | |
| NUDOCS 9208030010 | |
| Download: ML18058A986 (19) | |
Text
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Docket No. 50-255 Consumers ~ower Company ATTN:
Gerald B. Slade-
.. General Manager.
JUL 2 7 1SS2.
Palisade~ Nuclear Generatirig Plant 27780 Blue Star Me~orial Highw~y Covert, MI 49043
Dear Mr. Slade:
This refers to the inspection conducted by Messrs. J. K. Heller, Z. Falevits; J. F. Schapker, and others of this office on June 9 throug~ July 13, 1992.
The inspection includ~d a.review of authorized activities for your Palisades Nuclear Generating Plant.
At the conclusion of the inspection, the findings were*
discussed with those members of your staff identified in the ehclosed report.
The enclosed copy of our inspection report identifies areas examined during the
- inspection.
Within these areas, the inspection consisted of a selective examination of prbcedures and representative records, observations, and interviews with personnel. *
- During this inspection, *your activities pertaining to installation. of.the
~~alified incore detectors, as described in paragraph 8.b of this report, appear to be* in violation of NRC requirements. *However*, as described _in the report, you identified this violation.
Therefore, the violation will not be subject to enforcement action because your efforts in i derit ifyi ng and correcting the violation meet the criteria specified in Section V.G qf the "General Statement of Pol icy and Procedure for NRC Enforcenient Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C, (1991)).*
In accordance with 10 CFR 2.790 oi the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed. in the NRC Public Document Room.
We will gladly discuss any questiOns you have concerning this *inspection.
Sincerely, A' '
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B. L.
Jorgensen~ Chief Reactor Projects Section 2A
Enclosure:
Insp~ction Report No. 50-255/92018(DRP}
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- ~ Consumers Power Company Distribution:
REGION Ill 799 ROOSEVELT ROAD t
. GLEN ELLYN, ILLINOIS. 60137 i.,
Docket No. 50-255 Consumers Power Company *
ATTN:
Gerald General Manager
- Palisad~s Nuclear Generating Plant 27780 Blue Star Memorial Highway Covert, MI 49043
Dear Mr. Slade:
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This refers to the.inspection conducted by Messrs. J. K. Heller, Z. Falevits,
- J. F. Schapker, and others of this office on June. 9 through July 13; 1992.
The inspection included a review of authorized acti~ities for your Palisades Nuclear Generating Plant.
At the conclusion of the inspection, the findings were discussed with those members of your.staff identified in the enclosed report.
The enclosed copy of our inspectio~ report identifies areas ex~rnined duri~g the inspection.
Within these. areas, the inspection consisted of a selective examination of pr6cedures and representative records, obser~ations, and intervtews with personnel.
During this inspection, your activities pertaining to installation of the qualified incore detectors, as described in paragraph 8.b of this report,
- appear to be in violatio~ of NRC requirements.
However, as described in the.
report, you identifi~d this violation... Therefore, the violation will not be subject to enforcement action because your.efforts in identifying and ~prrecting the violation meet the criteria. specified.in Section V.G of th~ "General Statement of Policy and P~ocedure for NRC Enforcement Actions~w (Enforce~ent Policy, 10 CFR Part 2, Appendix C,.(1991)).
In accor.dance with 10 CFR. 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be pl aced in the NRC Public Document Room.
We will gladly discuss any questions you have concerning this inspection. *
Enclosure:. Inspection Report No. 50-255/92018(DRP)
Se~ Attached Distribution*
Sincerely,
- ~-
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B. L.
or e~, Chief Reactor Projects Section 2A y
U. S. NUCLEAR REGULATORY COMMISSION REGION II I Report No.
50-255/92018(DR~)
Docket No. 50-255 License No. DPR-20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson 1 MI 49201 Facility Name:
Pal~sades Nuclear Generating Plant I~spection At:* Palisades Site, Covert, MI Inspection Conducted:
June 9 through July 13, 1992 Inspectors:
J. K. Heller
- a. G. Passehl J. F. Schapker Z. Falevits.
- B. L. Jorgensen W. D. Shafer S. Sanders By:~~~ief.
..
Reactor Projects* Section 2A Approved Inspection Summary Date
.
Inspection from*June 9 through July 13, 1992 (Report.No. 50-255/92018(DRP))
Areas Inspected:
Routi~e unannounced inspectioh by the resident inspectors of action~ on pre~iously identified. items,* plant operations, react6r trips, radiolugical controls, maintenance,. surveillance, reportable events, and NRC Region III* requests.*
No Safety* Issues Management System (SIMS) items were reviewed.
Results: No unresolved items or deviations were identified. Two open items ~nd one non-ciied viol~tion were identified.
The strengths, weaknesses and open items are disctissed in par~graph.11,
"Management Interview."
In summary:
Strengths were noted. during post trip response, startup Plant Review Committee activities, manufacturirig of spent fuel dry casks, planning and proposed response for potential degradation of a primary coolant pump seal and supervi*sion of train~es during reactivity chahges.
Weaknesses were noted in a 10 CFR 50.72 notification, post trip review and a.
radiological shielding evaluation.
The non-cited violation is discussed in paragraph 8.b.
-*
9208030015 920727
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. PDR ADOCK 05000255 G
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DETAILS 1.
Pers6ns Cont~cted Consumers Power Company
- G~ B. Slade, Plant General Manager
- T. J. Palmisano, Plant Operations Manager P. M. Donnelly, Safety & Licensing Director
- K. M. Haas, Radiological Services Manager J. L. Hanson, Operations Superintendent R. B. Kasper, Maintenance Manager
.
K. E. Osborne, System Engineering Manager D.. J. Malone, Radiological Service Superintendent*
D. G.
Malone~ Operations Staff Support Supervisor K. A. Toner, Electrical/l&C/Comp~ter Engineering Manager
- R. W. Smedley, Licensing Engineer
- J. Haumersen, Electric~l/l&C Supefintendent
- J. 'L. Kuemin, Licensing Administrator Nuclear Regulatory Commission CNRC)
C. J. Paperiello, Deputy Regional Administrator W. D. Shafer, Chief, Reactor Projects Branch 2 B. L. Jorgensen, Chief, Reactor Projects Section 2A
- J. K. ~eller~ Senior Resident Inspector J. F. Schapker, Senior Project Inspector Zelig Falevits, Senior Project Inspector D. Passehl, Resident Inspectof
- Denotes some of those present at the M~nagement Intervi~w on July 15, 1:992 Other ~embers of the plant staff, ~nd member~ of the contract security force, were also contacted d~ring the inspe~tion period.
2.
Operational Safety Verification (71707, 71710, 93702, 42700)
Routine facility operating activities were observed as conducted in the plant and from the main control room.
Plant startup, steady state power operation and transient response were observed as applicable.
The performance of reactor operators and senior reactor operators, 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 coritrol room activities.
Evaluation, corrective action, and response for off nor~al conditions were examined.
This included compliance to any reporting requirements.
v
a.
General The plant operated at essentially full power during this reporting period, except as noted in paragraph 2.b.-
- b.
Reactor Trip On July 1, at 12:32 p~m, the unit tripped from 100 percent power, due to a loss of load, when the turbine monitoring/control ~omputers malfunctioned." The SRI and the Palisades Region III section chief; who was onsite for his quarterly site visit, responded to the control room to assess control room activities.
No p~oblems were noted with co~mand and control of the ~vent.
The malfunction occurred because connectors from several. circuit boards vibrated loose and generated the turbine trip signal. * A visual inspection: of the circuits by the vendor of the control system determined that the connectors were not properly secured following work activities during the last refueling outage.
Tests confirmed the loose connector theo~y.
Prior to startup the connectors were tested. to assure they were properly secured.
All safety systems responded as designed.
On July 3, at B:l7 p.m., the reactor was made critical ~nd the ~nit was returned to service at 2:23 a.~. on July 4.
During the trip; several components did not respond as anticipated.
(1)
The bA" non-safety related 4160 V bus did not fast transfer to an alternate power supply.
This occurred because. the alternate supply breaker had a mechanical interlock that was slightly out of adjustment.
Adjustments were made and the fast transfer circuit tested several times before ~eturning th~ plant to service.. This problem was not o~se~ved on any of the other breakers.
This blis is the power.source for two of the four primary c6olant pumps. Since this transfer occurred at approximately the same time as the loss of load, the crew initially diagnosed that the trip was caused from low primary coolant system flow.
The initial 10 CFR 50.72 notification - made at 1:00 p.m. - identified this as the cause of the reactor trip.
-This information was corrected during a
subsequent notification, approximately - three hours later when the inspector identified that the. initial
CFR 50.72 notification had not been updated.
(2)
The "B" safety injection tank depressurized by approximately 25 psi following the trip because of a relief valve that malfunctioned. It was repaired prior to returning the unit to service.
-**
(3)
(4)
Several *weeks earlier the safety. injection tank had depressurized by a similar amount while cover gas pressure was being adjusted.
The licensee conc1uded that the most likely cause was a relief valve problem th~t required an outage to facilitate a repair.
The repair was def erred to the next forced outage because the problem did not create a safety problem or affect tank operability.
The conttol ~com arinunciator chime malfunctioned for several min~tes. This problem has occurred on previous trips.
The problem did not hinder the operators ability to respond. This is the subject of a future design change.
- During the post trip review, one of the pl ant computers indicated that the upper detector for the "C" power range instrument detected high nuclear power for several minutes after the trip.
The problem was traced to a circlJit card, which was replaced prior to returning the unit to ~ervice.
The problem was isolated to the input to the computer and did not affect any of the inputs to the reactor protective system.
.
.
This item was of note to the-inspector, since the same problem was identified by the inspector during a review of.a December*
1991 plant trip.
At that time, the licensee indicated that
- it was a data display problem. The problem apparently was not identified for resolution during the recently completed 1992 ref~eling outage.
The inspector discussed this item wit~ several shift
.supervisors.
One supervisor indicated that this problem has occurred several times before.
A review of previous trip reports revealed that this has occurred during every trip for sever a 1.fuel eye 1 es. The inspector discussed this :item at the exit.
This problem did not create a safety pro bl em, but indicated a weakness with post trip data review~
.
.
(5)
Tours of the containment identified $everal minor leaks, that*
.were resolved without a mode change to facilitate repairs.
The inspector has no questions at this time.
Additional reviews will be performed when the LER is issued.*
c.
Plant Review Committee <PRC)
The inspector attended the startup PRC.
The it~ms documented above were presented by knowledgeable individuals who described the problems and the corrective actions. Jhe PRC membership discussed the problems and the technical merit of the solutions.
- The inspector verified that the PRC composition met Technical Speci fi cation composition requirements and a vet i ng quorum was
. present.
d.
- Criticality The inspector observed the licensee make the unit critical on July 3.
The cri ti cal ity was accomplished by diluting to predicted critical boron concentration.and then withdrawing the control rods to achieve plant criticality. The estimated critical rod height an~ -
boron concentration were within the predicted target band.
The reactivity changes were performed by train*ees. under *direct supervision of a reactor operator~
Additionally, a superintendent was on shift as a reactivity manager.
- e.
Tours (1)
Tours of the control room were routinely made. * During these visits the inspector. observed that shift personnel.
requirements were always met,. that the operators were cognizant of changing plant conditions, the equipment status*
board and.LCO board were maintained up-to-date and the operators were performing assigned tasks in accordance with plant procedures. Several of the activities observed were:
(2)
(a)
Co~trol rod movement per ~OP 6.
(b)
A mode change from hot shutdown to critical per GOP 3.
(c)
Power escalation after synchroniz~tion per GOP 5.
(d)
Post Trip Actions per EOP 1.
(e)
Reactor Trip Recovery per EOP 2.
.
(f).
Addition of cover gas to the Safety Injection tanks per*
SOP 3.
-
Tours of the auxiliary and turbine building were routinely performed.
Most were performed without the presence of the
- licensee staff.
Minor observatioris were* identified and resolved.
Several* observations pertaining to containment penetration cooling for the main steam lines were discussed ~ith system engineering.
(a)
The support for the duct work* for pen~tration 3, "Main Steam Line for S/G 8," was not attached to the wall~.
(b)
The support for the duct work for penetration 4, "Main.
Steam Line for S/G A," was a wire that had a rusty Jppearance..
(c)
B6th steam pipes were not centered in the penetration,.
which means that the ventilation flow nozzles were not evenly placed around the piping. Both penetrations have a temperature element that was located near ventilation
- nozzles.
The placement of the temperature probe does not appear to be in the optim~l position to obtain.
critical temperature profiles of the concrete nearest the piping.
These item~ were discussed with the responsible system engineering section chief.
These ebservations will require some review by the system engineer and are considered an open item until the reviews are 'complete.
(Open Item 255/92018-01 (DRP)).
f.
Primary Coolarit Pump Seal Staging On July 10, 1992, the lower seal for the "B" reactor coolant pump started to "un~stage" or malfunction.
The licensee evaluated the problem, established a forced outage schedule, determined that the lower.seal had not totally "un-staged. and the remaining seals were
"staging*i properly.
The pump design has *three seals that are capable of functioning with full system pressure and a fourth vapor seal. *.
The inspector reviewed alarm* response procedure number 5 and interviewed several operators. The operators were knowledgeable of the condition and able to discuss the symptoms of a failed seal.
The alarm respon~e procedure documented the operator actions and provided.the expected sea 1 1 eakoff pressure,. temperature and fl ow conditions if a single or several pump seals failed.
One open item and no violations, deviatiOns or unresolved items were identified.
- 3.
Radiological Controls (71707)
The inspector reviewed the Adi ve Pl ant Shi e 1 ding Log* and found three*
active shielding locations.
The inspector randomly selected shielding file 63 revision.I, "Shielding For Safety Injection, Containment Spray and Shutdown Cooling," for additional review.
The reviewwas accomplished by performing a visual inspection of the shielded area and using Administrative Procedures 7.14,. "Control and Use of Shieldjng and Associated Equipment"; 9.13, "Temporary Modification Control"; and 3.07,
"Safety Evaluations" as references.
a.
The inspector found that the *shielding consisted of six lead blankets firmly affixed around the
~ipe.
This was actually documented.in the file.
b.
The file contained quarterly surveys of the area.. Independent surveys performed by the inspector were in agreement with licensee results.
c.
The shielding file indicates that the shielding was "temporary" until the licensee was ready to flush the lines.
The shielding evaluation was dated September 1988. Administrative procedure 7;14
v
d.
e.
f.
does not define "temporary". However administrative procedure 9.13 does contain a definition of 90 days.
The shielding installation/removal record at Section 3 (item 2),
stated that the system was not in service.* This response eliminated the.requirement for a* safety.evaluation and only required that a static evaluation be considered.
The file contained *a one page calculation, showing that installation of three lead blankets was
~cceptable. This analysis was revised, dated October 1988, to state that the lines may be in service with six lead blankets.
What is not clear from the evaluation, is if the a.ccident dynamic conditions were considered:
The shieldirig evaluation was not processed as a Temporary Modification. Administration Procedure 9.13 does not specifically require a Temporary Modification for shielding but does imply one is.
required. If a Temporary Modification had been processed, then the safety evaluation*would have been performed and management attention would have been directed to the.age of the shielding due to the age of the temporary modification.
The inspector discussed this item with the Radiological Service Superintendent, who stated that -a program was underway to convert
- the active shielding items to Design Changes.
The expected completion date is Detembe~ 1992.
- The in~pecto~ identifi~d this item.at the end of the inspection period arid considers this an open item pending additional review by the inspector and the licensee to determine.if a Safety Evaluation is requi.red, a dynamic stress package is appropriate, and if this item should be addressed by the Temporary Modi fi cat'i on process.
(Open Item 255/92018-02 (DRP)).
One open item and* no. violations,. deviations* or unresolved items were*
identified.
4.
Maintenance (62703, 42700)
Maintenance acti*vities in the plant were routinely inspe~ted, iricluding both corrective maintenance (re~airs) and *preventive m~iritenance.
Mechanical, electrical, and instrument and control group maintenance activities were included as available.
The focus of the inspection was to as.sure the maintenance acti vit i e~
.reviewed were conducted in accordance with approved procedures, regulatory guides and. industry codes or standards and in conformance with Technical Specifications.
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 usirig approved procedures; and post
- maintenance testing was performed as applicable.
The licensee has revised the Corporate Quality Assurance manual to delete
v
the use of quality control i~spectors for verification of the work quality performed. by the p 1 ant maintenance department.
The qua 1 ity contro 1 * *
inspections have been rep 1 aced by inspections performed by maintenance personnel who have demonstrated an equivalent skill (or greater) of the individual performing the *repair acti_vities.
To monitor the implemeritation of the Quality Verification.Program (QVP) the li~ensee has*
established a temporary (approximately ?months) program that will monitor ahd document the progress of the QVP.
The QVP was placed into service on July 1.
The following work order (WO) activities were inspected:
a.
"Rebuild the P~568 Boric Acid Pump."
b.
- The QVP was impl.emented for.this activity. The inspector verified,
- by interview with the job supervisor and review of the WO, that the established quality verifications were performed by individuals with.
the appropriate skill. level and independence.
The job supervisor was able to discuss the QVP process and describe the independence and/or skill level required for the verification.
WO 24202281, "Repair Relief Valve RV-3128. For the "B" Safety Injection Tank."
The activity required the dedication of commerci~l grade parts to safety related service.
The inspector did a. cursory review of the
- dedication package and interviewed the procurement engineer.
The interviews indicated *that the engin.eer was knowledgeable of the dedication process.
- No violationsi deviations, unresolved or open jtems ~ere identified.
5.
Surveillance (61726, 42700)
The inspector reviewed Technical Specifications required surveillance
- .testing as described below and verified that testing was performed* in accordance with adequate procedures. Additionally, test instrumentation was calibrated, Limiting Conditions for Operation were met, removal and
- restoration of the affected components were properly accomp 1 i shed, and test results conformed with Techni~al Specifications and procedure requirements.
The results were reviewed by personne 1 other than the individual directing the test and deficiencies identified during_ the testing were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected or reviewed:
.*a.
Q0-19,
"Low Pressure Safety Injection Pump Test."
b.
Q0-"17,
"Service Water Pump Test."
.,
c.
d.
The inspector observed that the inservice inspection pump discharge pressure gauge (PS-1346) *was cycling approximately 8 *psi which required an estimate of the reading. The inspector noted that there was no attempt to use an instrument isolation valve to dampen the oscillations.
This was discussed with the operator who noted the observation as ~n item on the procedure improvement sheet.
Sl-7,
"Functional Test of the Fire Detection System Outside Containment."
The inspector reviewed Administr~tive Prricedure 10.41, "Procedure on Procedures."
Section 15.0 discusses the periodic review.process for procedures.* Section 15.0 specified a technical review every two years. The inspector questioned if this was the appropriate review cycle for refueling frequency surveillance t~sts. *If the fuel cycle is short, then the procedures could be reviewed and revised every other fuel cycle. *This could create a condition where lessons-learned are not incorporated until the.next fuel cycle.
This was discussed at the exjt interview.
.
.
No violations, deviattons, unresolved or open items were identified.
6.
Inspection of the ~entilated Con~rete Cask fabrication (37700)
a.
Background The 1 i censee contracted with Pati fi c Sierra Nuclear Corporation (PSN) to design and construct a dry cask spent fuel storage facility for long term temporary storage of spent fuel.
The licensee will document a 10 CFR 50.59 evaluation as required by 10 CFR 72.212 (Subpart K), demonstrating that use for dry stotage of spent fuel will not create an unreviewed safety question or require a Technical Specification (TS) change.
The PSN cask-design consists of a steel multi-assembly basket (MSB) *
which holds 24 spent fuel assemblies (sealed) and. a steel clad ventilated concrete cask (VCC) ~hich provides biological shielding and MSB protection.
PSN has NRC approval to build eight concrete casks and three multi-assembly baskets.
Th~ certificate of*
compliance that would permit use of the casks was pending NRC approval at the time of inspection.*
b.
Inspection This inspection wa*s conducted using the specifications, drawings, standards, codes~ and commitments described/in the ven~or's request for desigh certification. The inspector obse~ved the placement of concrete for VCC casks 4, 7 and 8.
The observations documented below apply to cask 8.
Additionally, the observations marked with an asterisk apply to casks 4 and 7.
( 1)
.Cement Type I I in accordance with ASTM C 150
y
"Specification for Portland Cement" was used.
(2)
Aggregate size and type was inspected and complied with the PSN Specification requirement.
(3)
The Concrete mix design comp 1 i ed with American Co.ncrete Institute (ACI)-318, Chapter 4 and 5, with compressive strength of 4000 PSI and course aggregate size 57, *speciffed.
(4)
Pla~ement of reinforce~ent was in accordance with drawing and specification requirements.
(5)
Preplacement inspection of concrete forms and material~ was made as required by specification requirements.
(6)
(7)
(8)
(9)
(10)
( 11)
(12)
Placement of the concrete in the forms was observed. and tomplied with ACI 301 requirements.
Observations of* the use of vibrators *to consolidate the concrete were in compliance with ACI 301 requirem~nts. *
Observations of inspections were requirements.
slump, temperature, and air content in
.compliance with applicable ACI
. Test cylinders were prepared in accordance with the American Society. for Testing Materials (ASTM) C31.
Testing.was performed by a independent testing laboratory with certified testing personnel utilizing* calibrated testing equipment.
All testing and placement ~f the concrete was inspected in progr~ss by the PSN inspector and the licensee's surveillance engineer.
Documentation of the fabrication and inspection criteria was
=
recorded on the process control sheet as required by the Specification (CVCC-89-001).
c.
Conclusion The licensee's corrective action taken in response to the unresolved item identified in NRC Inspection Report No. 50-255/92012(DRS),
addressed the weaknesses identified (Unresolved Item 255/92012-11).
The unresolved item remains open, however, until receipt and review by the NRC of the written response that was requested from the 1 i censee. *Test results for the tests performed by the independent laboratory have not been reviewed by the NRC inspector, but will be inc 1 uded with the QA Data Package when the 1 i censee accepts the vessels. The licensee plans to perform receipt inspection; and will review the ind~pendent labriratory results at that time.
NRC review
.
I
.
of the inspettion and fabrication records will be performed subsequent to the licensee's review.
- 7.
Safety Assessment/Quality Verificati6n (37701, 38702, 40704, 92720)
- The effectiveness of management contro 1 s, verification* and oversight
- activities, in.the conduct of jobs observed during this inspe~tion, was evaluated.
The i nspedor frequently attended management and supervisory meetings involving plant status-and plans and focused on proper coordination among departments.
.
.
The results of 1 icensee *auditing and correetive action programs. were routinely monitored by attendance at Corrective Action -Review B.oards (CARB) and by review of Condition R~port~. Problem Reports, Radiol~gical Deficiency Reports, and security incident report~.
- As applicable; corrective acti.on program documents were forwarded to NRC Region III technical specialists for informa~ion,and possible followup evaluation.
No violations; de~iations, uhresolved or open ~terns were identified.
8.
Reportable Events (92700, 92720)
The inspector reviewed the. following Licensee Event Reports (LERs) by means of direct observation, discussions with 1 icensee personnel, and review of.records.
The review.addressed. complianc~. to reporting requirements and,* as applicable, that immediate corrective action and
. appropriate action to pre~ent -recurrence had been accomplished.
.
a, (Closed) LER 91009: Qualified Core Exit Thermocouple Inoperable and Cannot Be Repaired While the Plant is ~t Po~er.
- ThiS informational LER was* submitted as a special* report after environmentally qualified coreexit thermocouple (CET) No. 16 w-as.
declared inoperable, and could not be repaired while the plant was at power.
The LER was issued per the reporting requirement of proposed Technical Specification (TS) change,. dated September _2, 1988.
The proposed TS required the restoration of the inoperable channel(s) within 7 days or the submittal of a special report to the NRC within 30 days, when the number of operable qualified CETs per quadrant is less than four but greater than or equal to two, The cause of the CET failure was an open lead between the CET and the recorder.. The remaining three CETs in the affected* quadrant*
were operable at the time.
The CET Was replaced during the 1992
- refueling outage. *The inspector has no further concerns regarding this issue and this LER is closed.
- b.
(Closed) LER 91011:
Seismic Qualification of Qualified Incore Detector Electrical Connectors.
I'
On June 27, 1991, the licensee determined that the environmentally qualified CETs were **in ari unanalyzed condition because their electrical connectors were not supported or connected as described
. in the* seismic analysis. The contractor performing head restoration work reported, after the plant was returned to service, that two supports were broken.. The contractor did not report the problem*
during the outage because he was not aware of the safety-significant application of the posts. Both NUREG-0737, "Instrumentation for the Detection of Inadequate Core Cooling", and Reg Guide 1.97 require that CfTs be environmentally and. seismically qualified.
- The inspectors documented their review of the licensee's interim operability justification for continued operation in Inspection Report 255/91012.
The. l.icensee subsequently determined that l&C
. technicians did not attach the CET cables to their supports as part of their installation sequence; therefore, the majority of t_he CETs were not supported as required.
The cause of the event was an inadequate procedure.
The procedure used by" the l&C group did not include instructions for attaching the CET cables to their supports.
As corrective action, the licensee revised affected procedures to include the appropriate instructions.
The 1 i censee a 1 so revised the design of the supports to reduce susceptibility to damage. These revisions. were incorporated during the 1992 refueling outage. Certain of these activities appear to be in violation of NRC require~ents. However, the licensee identified this violation and it is not being cited because the criteria
. specified in* Section V.G of the "General Statement of Policy and
- Procedures for NRC Enforcement Actions," (Enforcement* Pol icy, 10 CFR P~rt 2 Appendix C (1991)), were satisfied~ This LER is closed.
c.
(Closed) LERs 91-18 and 92-28: Electrical Cable Routing Anomalies.
(Closed) LERs 92-05, 92-07 ind 92-16: Lack Of Adetjuate Electrical Circuit Isolation.
An NRC electrical specialist was onsite.'on.July 15 and August 7, 1991, to evaluate licensee's electrical cable routing anomalies.
~eported in LER 91-14.
The inspection results were documented in report 255/91015.
- On June 10 and 11, 1992, a fo llowup inspection was performed to evaluate additional Appendix Rand other routing anomalies reported and to determine if Appendix R, "Fire.Protection" requirements were considered.
- Configuration Control Project (CCP) ptogram,.task 2.2.B, verified and upgraded the data base for the circuits and raceway schedule (CRS).
This program was established in_ May 1990.
Task 2.2.B reviewed the available cable routing design informati6n from schematic diagrams* and physical routing drawings of electrical cables/raceway.
y
'.
Additionally Task 2.2.B determined the appropriate channelization,*
physical routing, and implemented changes to the CRS data base.
Prior to the effort, 12,859 circuits had only 26 channel assignments and 9172 raceways had onlj 12 assigned channel data. The remaining CRS will be evaluated during this effort to confirm their existence and* to obtain the information necessary for channel assignments.
Task 2.2.B completiori date was scheduled for August 1992.
During this inspect~on, the inspector attempted to determine whether additional cable routing concerns~ spetifically those similar to the one reported in LER 92-028, existed.
LER 92-028, reported that the power supplies and cable routing for the Emergency Diesel Generator (EOG) rooms ventilation fans did not meet Appendix R requirements.
As a result, a single failure due to fire in the cable spreading room could render both EDGs inoperable *from elevated room temperature. due.to a *lack of adequate EOG roo~ ventilation..The.
licensee has instituted compensatory measures.which will be in effect until the cables are rerouted in the Fall of 1992.
.
.
.
.
The inspector reviewed LERs, engineering documents, design drawings,
- modification*s and interviewed engineering and operations personnel.
Based* on this review, the inspector "determined that the CRS anomalies* identified were mostly base*d on inconsistencies found in design documents. which were.verified by field walkdowns to confirm the potential conditions.
For* example, LER 91-014 identified.
documentation iriconsistencies for 38 safety-rel~ted cables. These in~onsistencies indicated that the cables were routed in the opposite (redundant) channel raceway which was contrary to cable
. separation requirements and FSAR commitments.
Subsequently, field
- routihg verifications, using a* cable electromagnetic signal inducer, determined that 101 additi~nal circuits were potentially ~isrouted~
.
.
.
During this process the licensee identified incorrectly channelized circuits, circuits routed in raceways other.than those the drawings indicated, and circuits routed in common* cable trays even though they performed redundant functions.
For example, six cable penetrati6ns containing 602 cables were found in the control room floor; these penetrations and cables had not been identified on existing drawings.
Partial field verifications of the 602 cables identified additional Class lE cables. Additional findings included circuits incorrectly classified as lE, raceways. not correctly chann~lized~ routing not as shown on drawings and modific~tions not entered in CRS.
As part of the 1 i censee' s proposed corrective action tb address the noted anomalies, some of the circuits noted above were determined. to require rerouting while others were dispositioned using engineering safety evaluations and an exemption request.
The licensee infor~ed the in~pector that the walkdowns have verified the *Critical circuits which were misrouted.. The cause of these deviations from separation requirements has been attributed* to (1)
a non-uniform interpretation of design criteria for channelizing
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'.
electrical cables, and (2) the unavailability of a single, complete and reliable source of documentation for electrical circuit routing and channelization.
Most deviations occurred during initial plant construction.
The inspector determined that Appendix R was not consi~ered in the scope of the CCP; however, the licensee was considering a Design Basis Documentation (DBD) review of Appendix R design data.
The licensee has initiated a pilot program to review and enhance*
Appendix R design documents and make the Appendix R program more user-friendly.
The licensee stated that completion-of the program
.will provide a much higher assurance and confidence that n_o significant safety issues exist relative to *violations of Appendix R separatiori requirements.
The inspector determined that whenever cable separation and channelization anomalies were identified, physical verification of in~talled routing was performed to determine adequacy of routing; cable ampacity, tray fill. and. if the installations met design requirements.
However, the inspector noted that the 1 i censee did not expand the. program to include. additional field walkdowns of
. safety-rel_ated cables that were not identified as discrepant during
- the de~~gn document reviews but could also have been potentially misrouted.
ND Violations, deviations, unresolved or open items were identified.
9..
Reqi on I II-Requests (92705)
a.
By memorandum dated June 19, 1992, Mr. J. G. Partlow, Associate Director of Projects, Office of Nucl.ear Reactor Regulation, requested data conterning the unavailabilitY of the emergency diesel generators due to testing or mainten~nce. (preventative or corrective).
The information was obtained by -reviewing the shift 1 ogs and the* di es el generators performance indicator 1 og.
The information was provided by separate correspondence dated July 9, 1992, to the Region III Technical Support Staff.
b.
On September 12, 1991, a utility discovered that a 450 pound Jet pump, stored in the spent fuel pool (SFP), dropped approximately 4,5 feet.
It was resting on fuel racks that contained five spent fuel assemblies; no fuel damage was observed. The pump was stored in the SFP since 1981 and secured to the side of the pool by a single
- .carbon steel cable. Surveys of the pool performed in 1990 andl991, indicated that the cable was showing signs of decay.
The cable was
. scheduled for replacement.
The NRC devel~ped a questionnaire to evaluate the licensee program for control of components, other thin spent fuel, that were stored in the spent fuel pool.
The inspector completed the questionnaire by interview with the onsite group responsible for foreign material
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..
exclusion control and a visual inspection of.the spent fuel pool~
The results were provided by separate correspondence dated July 15, to the Region III Technical Support Staff.
- No violations, deviations, unresolved or open items ~ere identified:-
10.
Open Items Open items are matters which hav~ been discussed with the licerisee, and will be reviewed further by the inspector.
These involve some action on the part of the NRC or licensee or both. Open items identified during the inspection are discussed in Paragraphs 2.~.(2) and 3.
- 11.
Mariaqement Interview The inspectors met*with licensee representatives - denoted in Paragraph 1
~ on July 15, 1992i to discuss the scope and findings of the inspection.
In addition, the likely information~l content of the inspection n~port with regard to documents or processes reviewed by the inspectors during the inspection was also discussed. The licensee did not identify any such
- * documents or processes as* proprietary..
A hon-cited violation was identified and discussed in the cover letter and paragraph 8.b, "Reportable Events - LER 9101°1: Seismic Qualification of Qualified Incore Detectors" of this report.
The non-cited violation pertained to the installation of incore detectors.
Highlights of the exit interview are discussed below:
a.
Strengths noted:
(1)
(2)
(3-)
(4)
(5)
Manigement of the response to the plant trip (pa~agraph 2.b,
"Reactor Trip").
Strong technical discussion of post trip report by the sta~tup PRC (paragraph 2.c, "Plant Review Committee (PRC)").
Use *Of trainees to perform ~eactivity changes and the close supervision provided by the reactor operators (paragraph 2.d,
"Criticality").
Preplanning activities, operator training and degree of written response* procedure for a "un-staged" Primary Coolant
. Pump Seal (paragraph 2.f;
"Primary Coolant Pump Seal Staging").
Improvements made in the process for manufacturing of the concrete spent fuel cask (paragraph 6, "Inspection of the Ventilated Concrete Cask Fabrication").
b.
Weaknesses noted:*
v
.J (1)
(2)
(3)
Incorrect information provided. in..the I 0 CFR 50. 72 notification and the delay in resolving the error (paragraph 2.b.(l), "Reactor Trip").
Proble~ with post trip review of information and the failure to identify/resolve a recurring problem (paragraph 2.b.(4),
"Reactor Trip").
Number of administrative errors contained in a shielding evaluation (paragraph 3, "Radiolrigical Controls").
c.
Open Items The open items pertaining to containment penetration cooling (paragraph 2.e.(2), "Tours") and shielding evaluations (paragraph 3,
"Radiological Controls") weie discussed.
- d.
The general topic of Quality Verification was discussed.
The inspector stated that Riii was aware of the licensee implementation schedule and will look at the program during subsequent inspections (paragraph 4, "Maintenance").
- e.
The periodic procedure* review process and if program enhancement
. could be made as the. process was applied to -refueling-frequency surveillance procedures was discussed (par~graph 5.d,
"Surveillance")..
f..
The inspector questioned if shielding that was placed on piping did not classify as a Temporary Modification and require the controls of a Temporary Modification (paragraph 3.e, "Radiological Controls").
g.
- The two Region II I requests were di scu~sed (paragraph 9. a & b,
"Region III Requests").
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