IR 05000312/1987021
| ML20237G968 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/27/1987 |
| From: | Ang W, Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20237G963 | List: |
| References | |
| TASK-2.K.3.05, TASK-TM 50-312-87-21, GL-86-05, GL-86-5, IEB-85-003, IEB-85-3, NUDOCS 8708170005 | |
| Download: ML20237G968 (33) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report No.
50-312/87-21 Docket No.
50-312 License No.
DPR-54 i
I Licensee:
Sacramento Municipal Utility District (SMUD)
P. O. Box 15830 Sacramento, California 95813 i
Facility Name:
Rancho Seco Nuclear Generating Station l
Inspection Conducted:
June 15-25, 1987 Inspection by:
7 M~ U W. P Ang, Project nspector Date Signed
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Approved by:
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7'27-h
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i LGil'ler, ChigProject Section 2 Date Signed f
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i Accompanying Personnel:
D. A. Beckman, Prisuta-Beckman Associates l
G. Morris, Westec i
, Summary:
Inspection on June 15-25, 1987 (Report No. 50-312/87-21)
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Areas Inspected:
Routine announced inspection by a region based inspector of licensee action on previously identified inspector items, and Licensee Event i
Reports.
Inspection Procedures 30703, 36100, 92700, 92701, and 92702 were j
used during this inspection.
Two contractors accompanied and assisted the
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inspector with this inspection.
Results:
In the areas inspected, one violation regarding failure to perform j
wire and cable routing inspections was identified (paragraph 2.A).
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QB170005070730
ADOCK 05000312 PDR
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DETAILS 1.
Personnel Contacted l
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- B.' Croley, Director, Technical Services l
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- D. Keuter, Director, Operations and Maintenance
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S. Knight, Quality Assurance (QA) Manager
- G. Cranston, Nuclear Engineering Manager D. Army, Nuclear Maintenance Manager
- J. Vinquist, Licensing Manager (Acting)
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- R. Cherba, Quality Engineering Supervisor l
- P. Hypnar, Impell, Cable Raceway Tracking System (CRTS) Project Manager
- T. Fetterman, United Energy Services Corporation, Electrical Engineering
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Manager
- E. Gough, Bechtel, Nuclear Engineering Department, Supervisor
- J. Browning, Incident Analysis Group
- J. Hodges, Incident Analysis Group
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- J. Meyers, QA Auditor
- J. Robertson, Nuclear Licensing Engineer
- Attended the exit meeting.
The inspector also held discussions with other licensee and contract personnel during the inspection.
This included plant staff engineers, technicians, administrative and clerical assistants.
2.
Onsite Follow-up of Written Reports of Nonroute Events A.
Licensee Event Reports (LER) 85-16, 86-10, 87-13, 87-24, 87-26 -
Cable Raceway Tracking System (CRTS) Related Items The subject LERs reported various discrepancies related to cable
design and installation.
LERs 85-16 and 86-10 and the CRTS reported i
discrepancies had been previously inspected and the inspection scope l
and findings were documented in Inspection Report 50-312/87-06.
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On April 3, 1987, the licensee submitted to the NRC a Wire and Cable Program Description and Action Plan.
On May 1, 1987, Region V provided the licensee written comments regarding the plan.
On
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May 6, 1987, the licensee met with representatives from NRR and
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Region V to discuss the program.
As a result of the Region V and NRR comments, the Engineering Department Manager stated that a revision to the CRTS Program would be submitted to the NRC.
Representatives from NRR and Region V met with the licensee and I
performed an inspection between June 15, 1987 and June 25, 1987, to review the licensee's progress regarding resolution of the cable routing problems.
The licensee had identified types of concerns ranging from simple documentation problems to overfilled raceways,
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intermixing of cables, and errors in cable routing.
The inspectors l
noted that the majority of the misrouting concerns had occurred l
during the 1983 outage, when approximately 1100 safety-related cables were installed.
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The inspector, an NRC consultant and an NRR reviewer held discussions with the licensee's incident analysis group (IAG) to determine the licensee's current root cause analysis for the subject LERs.
It was noted that the root cause analysis was still being performed.
A review of the preliminary IAG activities indicated that a thorough, detailed evaluation was being performed.
Records l
and hardware for the subject discrepancies were being reviewed,
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applicable procedures and practices were being evaluated and l
interviews with available personnel were being performed.
In the review process, corrective action recommendations were.being i
developed although not yet finalized.
The NRC representatives
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informed the licensee that there appeared to be a need-to factor IAG findings and recommendations promptly into the ongoing Wire and
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Cable Corrective Action Program rather than waiting for the l
completion of the parallel root cause analysis effort.
The licensee j
acknowledged the need to do so.
l The NRC inspectors and reviewer witnessed a demonstration of the
current Cable Raceway Tracking System (CRTS) computer program i
application to understand the licensee's current methodology for j
managing the plant's cable population.
Discussions were held with
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the licensee's staff to understand the previous Bechtel system
(known as the EE 553 program) and the change over to, and development of CRTS.
The CRTS System does not automatically determine
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appropriate cable routes, but instead is a recordkeeping data base j
identifying what raceways the designers have selected for the j
individual cable routing.
LERs 86-10, 87-13, and Occurrence J
Description Reports (ODRs) 604 and 723 all pertain to cables found in incorrect raceways.
As an aid in confirming the routing of a selected sample of suspected cables, the licensee was using a DASAR current tracer which imposed a high frequency signal on the desired conductor.
This permitted the licensee to track the cable through the raceway system.
The NRC representatives observed the signal tracing of a sample cable, hand-over-hand tracing of a sample cable, and a partial cable pull through a cable tray for a new cable installation.
The signal tracing of the sample cable identified that it was routed through a conduit different than that documented in the CRTS.
This alternate conduit was the same separation group and was routed between the same two raceway termi%als through a different route.
This was judged to be a minor discrepancy by the licensee since the misrouted condition was acceptable as is, and, in that sense, had no safety significance.
The NRC inspector performed an inspection.to verify licensee compliance with USAR commitments and NRC requirements in relation to the licensee's implementation of its wire and cable corrective action program.
Rancho Seco USAR Sections 7.1 and 8.2 commit SMUD to various design and installation criteria for electrical equipment and cabling including redundant channel separation, power / control and instrument cabling separation, and cable tray fill criteria.
Rancho Seco USAR Appendix 1B refers to the operations Quality Assurance (QA) Program for QA commitments and requirements.
The QA
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Program policy Section 1 commits to Regulatory Guide 1.30 and ANSI N45.2.4-1972, " Installation, Inspection, and Testing Requirement for I
Instrumentation and Electric Equipment During the Construction of
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t& clear Power Generating Stations, for QA requirements for the
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installation, inspection and testing of instrumentation and electric
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equipment.
Prior to applicability of the current USAR section and j
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the operations QA Program commitments, on September 23, 1976, the
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licensee had enmmitted to a QA Program in accordance with WASH 1284,
" Guidance on QA Requirements During the Operations Phase of Nuclear Power Plants," October 26, 1973.
The licensee's commitments were confirmed by an NRC letter from R. Reid, DOR, to J. Mattimoe, SMUD, on December 13, 1976. WASH 1284, includes Safety Guide 30 for QA
requirements for the installation, inspection'and testing of j
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instrumentation and electric equipment.
Safety Guide 30 accepts h
q ANSI N45.2.4-1972 (IEEE STD 336-1971) for QA requirements.for the j
installation, inspection and testing of instrumentation and electric
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equipment.
ANSI.N45.2.4-1972, paragraph 5.1.1,' requires inspections-
to verify correctness of installation including inspections, as-
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I appropriate, for. proper location and routing of cables and sensing v
lines in accordance with latest approved for construction drawings.
The inspectors divided the scope of the Wire and Cable Corrective
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Action Program into the following categories to better define the j
discrepancies identified by the licensee:
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Original Construction Cable Installation
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l Current Cable Installation
.I 1975-1986 Cable Installation
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CRTS Data Base Discrepancies (1) Original Construction Cable Installation The licensee had reviewed the procedures and practices during
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plant construction to justify exclusion of the original i
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construction cable population (approximately 14,000 cables)
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from the licensee's sampling reinspection (signal tracing)
program.
The licensee concluded that the cables installed and inspected during original construction did not have to be re-inspected or sampled because the program in place and implemented during construction provided reasonable assurance i
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that the cables were installed and verified to be installed in accordance with construction drawings.
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The NRC inspector reviewed.the following:
Quality Control Instruction 204A, October 6, 1970, Installation Inspection of Electrical Cable and Wire Selected Construction Inspection Data Reports (CIDR) for Installation of Electrical Cable and Wire during original construction The above noted procedure and selected CIDRs appeared to be sufficiently detailed to provide reasonable assurance that
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quality control (OC) inspections verified installation and I
routing to be in accordance with the design.
j However, the licensee had still been unable to identify the
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specific location for all Class I cable pull inspect 3cn records l
generated during original construction.
The licensee was
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requested to obtain the original construction pull cards for the cable affected by LER 85-16, Spurious Closure of DHR Drop l
Line Isolation Valve, to determine what inspection verification
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was accomplished for those cables.
The cable installation and l
inspection prograin for original construction appeared to I
provide reasonable assurance of correct installation.
However,
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the inspector obst.rved that inspection records to verify
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corrt-t installation should be available.
The licensee ack oA edged 1.he inspector's comments.
(2) Current Cable Installation
The current procedure for cable installation and inspection at
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Rancho Seco is Modification Procedure / Inspection Standard (MP/IS) 307, Revision 0, " Cable Installation." The procedure was reviewed for compliance with ANSI N45.2.4 routing inspection requirement.
MP/IS 307 cont;ined more installation and inspection details than the procedures in use during the
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pre-1986 timo period. MP/IS 307 included a cable inspection
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record that had an inspection attribute for correct routing.
Observation of cable pulling activities nd discussions with QC inspectors indicated that currunt SMUD practice included.QC witnessing of Class 1 (safety relatedl cable pulling and QC verification of cable routing.
Howeve'r, MP/IS 307 is l
applicable to Quality Classes 1, 2, and 3 cable. ' Consequently,
MP/IS 307 QC inspection requirements were. frequently stated as
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"when required" or if " applicable".
The procedure did not
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state specifically when QC inspection is required or whether it l
was always applicable for Class 1 cables.
The licensee was informed that although current QC inspection practices included Class 1 cable routing inspection, improvement of MP/IS 307 to i
specifically incorporate the ANSI N45.2.4 required QC inspections cippeared necessary to prevent future misinterpretation.
The licensee acknowledged the inspector's comments.
(3) 1975 to 1986, Cable Installation
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The inspector reviewed the-historical record of routing errors in this period and the licentsee's diagnostic efforts to date.
LER 86-10 reported seven safe shutdown indicction circuits that
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were found unpoce.ected from fire in t.' single switchgear room.
The LER further stated that the cables were required to be installed in a conduit but was found to be routed in an unwrapped cable tray.
Ecploy'ee concerns regarding overfilled trays and incomplete CRTS data led to an enhancement program
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I for the CRTS to check for raceway overfills, violations of separation criteria, intermixing of instrument cables with i
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power and control cables and other discrepancies.
Numerous discrepancies were identified by this enhancement program and are discussed in a subsequent section of this report.
In addition, the licensee decided to " signal trace a significant sample" of the approximately 2400 safety cables installed from 1975 through 1986 and compare the as-built condition with plant
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documents and the CRTS data base to establish a level of i
i confidence in the reliability.of the data base.
The licensee j
has chosen a sampling program that was intended to demonstrate
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that there was 95 percent confidence that less than 5 percent of the plant's safety-related cables " violated plant safety" because of routing that differed from the CRTS data base.
The licensee's Wire and Cable Program submitted to the NRC on April l
l 3, 1987 defined the sample population and data as of that date
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as follows:
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Description Lot 1 Lot 2 Lot 3 Lot 4 Population size 424 1702 190-
l Sample Size (238*)
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78 l
Circuits traced
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76 l
l Major defects
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7 Minor defects
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Insignificant
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or no defects
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67 Definition of Lots Lot 1:
Safety-related and safe-shutdown circuits with rerouted vias between commercial operation and December 22, 1986.
Lot 2:
Safety-related and safe-shutdown circuits without rerouted vias between commercial operation and December 22, 1986.
Lot 3:
Safety related and safe-shutdown circuits without revised vias and with questionable cable pull card signatures.
Lot 4:
Safety-related and safe-shutdown circuits with rerouted vias and questionable cable pull card signatures.
- After the initial sample section, Lots 1, 2, and 3 were reformed to add Lot 4, and one additional circuit in Lot 2 was traced and three additional circuits in Lot 3 vere traced, but these circuits were not selected by the random
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sample.
The total number of circuits 'o be traced, t
therefore, equals 238.
As noted above, signal tracing of cable in Lot 4 identified seven additional misrouted cables (beyord those in LER 86-10)
that were subsequently reported in LER 87-13.
At the time of this NRC inspection, tracing of Lots 2, 3 and 4 had been completed.
No additional " major defects" were identified in those lots although " minor defects" had been identified in all lots.
However, the sampling of Lot 1 identified a major defect on the 52nd circuit sampled.
The original 52nd circuit could not be released by Nuclear Operations for signal tracing due to
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plant conditions.
The licensee selected a substitute 52nd sample and subsequently determined that it was a misrouted power cable installed in an instrument cable tray (Occurrence
Description Repurt (ODR) 604).
On June 16, 1987, during this
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NRC inspection, the licensee issued ODR 723 identifying nine additional misrouted cables associated with this defect.
ODR 723 resulted from the licensee's' appropriate action to check nine other' cable installations having similar characteristics (i.e., same field engineer, same timeframe, etc.) as the ODR 604 identified misrouted cable.
In light of the identified discrepancies in Lot 1, the licensee was considering 100 percent routing inspection of the rerouted portions of Lot I cable at the end of this inspection.
Discussions with the licensee and IAG indicated that the licensee was uncertain regarding the procedural requirements and inspections performed to verify cable routing in the 1975-1986 time. period.
The licensee's QA department informed the NRC inspector that no NCRs or ODRs had been written which l
identified that required QC inspection of cable routing had not been performed during the 1975-1986 timeframe.
The NRC inspector reviewed the following procedures and interviewed present and past QC electrical inspectors to determine the licensee's procedural requirements and actual practice
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regarding QC inspection of cable routing during the 1975-1986 time period:
Engineering and Inspection Instruction (EII) EC-11, Revision 6 (effective November 1, 1984), Cable Installation During Construction and Major Modification Quality Control Instruction (QCi) 107, Revision 4, Installation Inspection of Electrical Cable, Wire and Conduit The licensee's current QC lead electrical inspector informed the NRC inspector that 90 to 95 percent of cable installation inspee.tions during the 1975 to 1986 timeframe were performed in accordance with FII EC-11.
Paragrapn 7.5 of EI: EC-11 required the QCE (QC inspector) to verify cable routing in accordance with drawings.
EII EC-11 Figures 1 and 2 are cable routing cards where installation verification by the field engineer t
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i (FE) and QC was required to be documented.
The specific signoff spaces however signified " completed by" (the FE) and-
" inspected by" (the QCE).
The cable routing cards of EII EC-11 contained no signature space for specific acceptance by-QC of cable routing.
The current lead electrical QC. inspector informed the NRC
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inspector that some of the electrical QC inspectors during the i
1983 outage, who no longer are QC inspectors but are still working onsite, had informed him that verification of' cable routing had not been completely performed during the 1983 outage.
The three individuals, who were QC electrical inspectors during the 1983 outage, were interviewed by the NRC inspector.
During the interviews, the individuals informed the NRC inspector of the.following:
There was an average of three electrical QC. inspectors i
during the 1983 outage.
Each QC inspector was assigned approximately 500 cable installations to verify during the 1983 outage.
Fifty to 60 percent of the pull cards were received by the j
QC inspectors for inspection signoff after the cables had
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j already been pulled.
l Two of the three individuals stated that the lead
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electrical QC inspectcr verbally instructed them to
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inspect routing by inspecting as much cable as possible l
from its termination points (ends) without going through
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the trays since they could not afford the QC manpower to witness 100 percent of the cable routing.
(That previous lead electrical inspector was no longer on site and was not interviewed.)
All three individuals stated that.they seldom verii"ied routing by looking at all the trays.
In addition, when they looked at trays, one of the inspectors stated the teays were usually covered by the tira they performed their inspections.
Based on the information provided by IAG, Nuclear Engineering, QC and the three former QC inspectors, and, based on tie misrouted cables identified in LERs 86-10 and 87-13 and ODRs 604 and 723, the inspector concluded that the ANSI N45.2.4, paragraph 5.1.1, requirements for inspection to verify proper location and routing of cables were not performed during the 1983 outage.
The failure to perform these inspections is
considered a violation of 10 CFR 50 Appendix "B" Criterion X and was identified as Violaticn 50-312/87-21-01, failure to perform cable routing inspections.
The licensee indicated to the inspector that all misrouted cables id(ntified so far by signal tracir;g were misrouted
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1 during the 1983 outage and were involved in a series of design l
changes where those cable's routing was changed more than once i
during that outage.
As noted previously in this inspection l
report, because of the identification of misrouted cables documented on DDRs 604 and 723, the licensee was considering
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100 percent reinspection of the rerouted portions of Lot 1 cables (1975-1966 time period rerouted Class 1 cables) at the
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conclusion of this inspection.
Based on the information obtained during the inspection, the NRC inspector identified
concerns regarding the adequacy of sampling Lots 2 and 3, in lieu of performing the required inspections.
These concerns will be forwarded to NRR for resolution.
(4) CRTS Data Base Discrepancies I
As a result of the discrepancies identified in LERs 85-16 and 86-10 and employee allegations to the SMUD Ombudsman concerning cable tray overfills and the validity of the CRTS data base,
l the licensee initiated an enhancement program for the CRTS data
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base to check for raceway ovorfill;, violations of separation i
criteria, intermixing of instrumant cables with power and
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control cables and other discrepancies.
In July of 1986, SMUD contracted the Impell Corporation to perform a review and i
evaluation of the CRTS data for the plants total 23,000 cable population.
The review resulted in approximately 1800
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l discrepancies identified in Class I cable.
The discrepancies l
were reported on CRTS problem reports, NCRs, ODRs and LERs 87-24 (overweight cable trays) and 87-26 (mixing of I
power / control cable with instrumentation cable) depending or, significance.
During this NRC inspection, the adequacy of the licensee's evaluation and disposition of the identified CRTS data base discrepancies were reviewed.
The NRC review centered on the LER 87-24 reported issue regarding overfilled Class 1 cable trays and conduits, a.
The licensee justified overfilled raceways by a nurber of calculations.
The Region V representative reviewed the calculation which addressed overfilled Class 1 cable trays.
It did not appear that the additional cables result.:o in any overweight problems with the Class 1 trays.
However, this calculation, performed by the licensee's agent, contained the following apparent discrepancies in the calculation and justification of cable ampacities:
1)
The USAR required 125 percent ampacity chargin overload current was not included.
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The discrepancies included overfi?1 conditions in relation to the installation of battery charger supply cables.
These conditions were determined by the licensee's agent to be technically acceptable on
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the basis of calculation 271-101-102, Revision 1, May 4, 1987, page 8.
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l However, this calculation of the battery charger load
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l current only included the normal steady state de load
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and did not include the current required to recharge
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the battery.
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The acceptance values for cable ampacity were based i
upon SMUD's Cesign Guide 5204.24, " Cable Derating i
Practice," and ICEA Publication P-54-440, "Ampacities of Cables in Open Top Cable Trays." These values were consistent with the USAR cable tray fill limit criterion of 40 percent, and provided conservative ampacity values for higher cable' tray fills. When the NRC representative corrected the battery charger i
load current error noted above, the existing cable I
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would not meet the stated acceptance criteria by at I
least one cable size.
The licensee responded that these battery charger cables were part of the original plant design and their size was selected based upon the Bechtel-Rancho Seco design manual.
The cable ampacities given in this earlier design criteria were from ICEA Publication P-46-426, Power Cable Ampacities, Volume
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This earlier standard permitted higher
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ampacities.
However, it limited the number of power conductors that could be grouped together.
The licensee's engineering consultant stated that because Rancho Seco mixes power and control cables in the same cable tray, the origins 1 design was based on cable ampacities derated for groupings of 25 to 42 power conductors.
The RV representative compared the cable area of 42
conductors of different cable sizes with the area of l
a 12 X 3 inch cable tray and determined that the I
percent tray fill if the smaller power cables were used would equate to acceptable fill limits
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considerably below 40 percent.
The licensee could not provide any correlation between the number of power cables in the power / control cable trays and the cable tray limit of 40 percent.
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The Region V representatives also reviewed the calculation which addressed overfilled Class 1 conduits.
The justifications used in this calculation focused on the potential for physical damage during cable installation.
The actual current loading on the cables and the potential for overheating of the cables were not specifically
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addressed other than a statement included from one of the manufacturers of the cables installed at Rancho Seco who stated that an increase in percentage fill will improve cable ampacity.
This statement appeared to conflict with-
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industry practice and industry codes.
The. Region V i
representative attempted to verify this statement with the manufacturer but was not successful.
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Pending licensee's confirmation of the adequacy of cable
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ampacity in overfilled raceways, the above items were identified as unresolved item 50-312/87-21-02, cable tray / conduit overfill /ampacity questions.
(5) 0A/QC Involvement in Wire and Cable Corrective Action Progeso SMUD QA representation in the Wire and Cable Corrective Action Program was evident during this NRC inspection.
However, j
active involvement in the decision making process was only j
evidenced by the QA managers' approval of the March 31, 1987,
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Wire and Cable Program submitted to Region V.
The NRC
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inspector had the following observations and/or concerns I
regarding QA involvement in this program:
a.
The applicability of ANSI N45.2.4 (IEEE STD 336-1971)
ins;,ection requirements appeared to be a surprise to the
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licensee including QA representatives.
The inspector l
could not identify any reference to those standards in the I
current inspection procedure.
b.
The QA/QC department could not determine during the inspection whether an NCR or ODR was required for the lack
of QC inspection of routing during the 1983 outage despite
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their knowledge of the QC inspector practice of only inspecting the ends of the cable routing and despite several inquiries by the NRC inspector.
c.
SMUD QC appeared to be only required for cable determination and termination during the signal tracing.
Impell QA/QC presence during signal tracing and other Impell Wire and Cable Program activities were not readily apparent to the NRC inspector.
If the signal tracing is going to be considered by SMUD as the replacement for the required independent cable routing verification, the need to assure full' compliance with the current approved SMUD QA program '.tas emphasized by the NRC inspector.
This should include adequate inspection independence, independent design reviews, inspection records, and records storage / availability.
d.
The QA/QC department had previously been aware of most of the facts that resulted in the NRC inspector concerns in paragraph (3) above and responded to the NRC questions with statements such as "That's a good question."
However, active QA involvement was not evident in pursuing
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f resolution of most of those questions when they became I
aware of those facts (for example, lack of QC routing (
inspections during 1983 outage).
I Pending completion of the licensee's Wire and Cable Corrective
Action Program, approval of the program by NRR, and resolution of the NRC inspector's concerns; LERs 85-16, 86-10, 87-13, 87-24 and
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87-26 were left open.
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LER 86-15-LO (Clostd) - RM-80 Printed Circuit Board Workmanship l
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This item was previously inspected during Inspection 50-312/87-13 and was left open pending further NRC review, i
The licensee reported that, during cold shutdown conditions, two trcce solder pads were dislodged from a printed circuit board during
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repair of the radiation monitor computer communications board for Radiation Monitor R-15050.
The report indicated that the failure was caused by poor soldering techniques used by the licensee's technician.
l The previous inspection concluded that additional review was l
required by NRC to determine whether the affected component was a i
" basic component" as defined by and subject to 10 CFR 21 and whether an additional licensee 10 CFR 21 report should have been filed.
The matter was discussed with regional inspection specialists familiar with the specific equipment and reporting requirements.
The inspector concluded that the circumstances were not subject to further reporting per 10 CFR 21, and that the circuit board was not i
a basic comoonent.
Further review onsite during this inspection
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determined that no additional similar failures have occurred.
This item is closed.
No violations or deviations were identified.
l C.
LER 85-12-L1/L2 (Closed) - Loss of Reactor Building Integrity l
Revisions 0 and 1 cf the LER and the licensee's corrective and l
preventive action plans were reviewed during Inspection 50-312/87-16
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(May 4-8, 1987).
During that inspection, the licensee's plans and actions, as documented in the above LER revisions, were determined to be inadequate and untimely to correct and prevent recurrence of airlock door malfunctions causing the losses of reactor building integrity.
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At the meetings with plant management during the prior inspection, the inspectors were advised of more extensive licensee plans which were not included in LER Revisions 0 and 1.
These included contracting the containment airlock vendor to train licensee personnel on vendor experi.ence with the doors, a maintenance tuneup on the doors, and a vendor review and recommendations for door improvements and maintenanc.
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Based on those discussions with plant management, the inspectors requested that the LER be further revised by the licensee to provide commitments to NRC for the more extensive actions documented above.
After concurrence by the licensee's Chief Executive Officer, the Plant Manager advised NRC Region V (subsequent to Inspection 87-16)
that such a revised report would be issued.
LER 85-12, Revision 2, was issued on June 12, 1987, adequately documenting the licensee's plans as commitments to NRC.
Revision 2 was revicwed and found acceptable for conformance to prior discussions and the results of Inspection 87-16.
This LER is closed.
No violations or deviations were identified.
D.
LER 85-02-LO (Closed) - Unrestrained _ Heavy Load in Reactor Building On January 28, 1985, during power operation, the. licensee discovered that two components of the Self Powered Neutron Detector (SPND)
disposal equipment were improperly st "ed in the' reactor building.
l The 12,000 lb. SPND basket cask was c ared at the +60 ft building
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level rather than at the drawing specified 15 ft level of the Fuel Transfer Canal.
The 10,700 lb. SPND transfer cask was stored on top of the stairwell at the +70 ft level in a horizontal position rather than in the drawing specified incore service area in a vertical
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l position.
Licensee analysis determined that the unrestrained transfer cask load was seismically unstable and capable of impacting the containment spray header during a seismic event; the basket cask load was determined to be acceptable.
The licensee further i
determined that the event was caused by the failure to recognize and correct a deficient storage procedure.
The procedure instructions l
were ambiguous and had contributed to improper placement of the loads.
The inspector reviewed the LER, determining that it was issued in a timely manner and included the required information.
Specific licensee corrective and preventive actions were verified as follows:
Records of initial personnel retraining on March 14 and 18, 1985, were reviewed and found to include personnel attendance records and subject matter germane to the incident.
The licensee has planned recurring training for cognizant personnel to be conducted prior to each refueling. outage.
Classroom Training Outline MM21P1301, " Rigging, Handling, and Crane Operation, Reactor Building Hoisting," includes training responsive to prevention of similar incidents and specifically includes review of the subject and similar events.
The Training Department "WorkTrak" computer system was verified to include a recurrent tracking item (Item #870735) to identify j
the need for and schedule training per the licensee's j
commitments.
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The inspector reviewed the following procedure revisions to confirm that the licensee has corrected the ambiguous procedures and included new provisions for verification of proper storage conditions prior to~ reactor building closeout.
M-135, Control of Heavy Loads, Revision 1 M-30, Incore Monitor Handling and Disposal, Revision 13
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AP-20, Reactor Building Closeout, Revision 4 l
Using these procedures, the reactor building was inspected by the I
NRC inspector to confirm that the SPND loads and other similar heavy
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loads were stored in accordance with the procedural requirements.
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No discrepancies were identified with heavy loads.
However, numerous unsecured construction items were observed and discussed
with the licensee's building construction coordinator.
The j
coordinator was cognizant of these conditions and plans were in
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place to clear the building prior to plant restart.
This LER is i
closed.
No violations or deviations were identified.
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E.
Control Room Essential HVAC Events The following LERs each involve problems occurring with the Control Room Essential HVAC (CR HVAC) System.
The preventive and corrective
actions for each event have been included in the licensee's System l
Review and Test Program (SRTP).
These actions include a commitment i
to correct the identified discrepancies prior to restart.
As l
indicated below, some of the actions have been completed; others
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remain in progress.
NRC is routinely reviewing the status of, and inspecting the implementation of the SRTP.
Based on correlation
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between the LERs and the SRTP established during this inspection, the following LERs are closed.
(1) LER CB-07-L1 (Closed) - Control Room Emergency HVAC High Flow Condition
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The report documents discovery of system high flows found during routine testing which could have reduced accident filter retention time below values required by safety analyses.
The problems identified included discrepancies between Interim Data Acquisition and Display System (IDADS) indicated flows and directly measured flows, differences between makeup and recirculated air flow rates, and unacceptable flow stabilization after-start times.
The above items are identified as SRTP Problem Tracking Items 26.0005, 26.007, and 26.0010.
As of this inspection, all planned modifications to the CR HVAC system have been completed and testing per STPs 779A and B was in progress to resolve continuing flow imbalance problems.
This testing was being followed by the NRC resident inspectors and Ea NRC c7ntd1 tent.
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(2) LER-86-09-L1 (Closed) - Spurious CR HVAC and Diesel Generator
'A' Actuation l
While ir cold shutdown, personnel lifting leads for a clearance
tag boundary inadvertently shorted the lead, causing a spurious
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automatic start of the "A" Diesel Generator.
No genuine ESF
signal was present.
Corrective actions were to realign the
Diesel Generator for normal operation.
l In an unrelated event, the
"A" CR HVAC System started from a spurious high area temperature signal while the system was considered out of service.
The LER indicated that the high j
temperature signal was believed to be due to installation of j
the sensor in an area which was not representative of the bulk I
area air temperatures in the control room and technical support I
center areas served by the system and the matter would be I
resolved as part of the SRTP.
This problem is similar to
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control room ventilation problems described by.IE Information
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Notices 85-89 and 86-76.
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SRTP Problem Tracking Item 26.0018 documented the problem.
The
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inspector reviewed the results of a temperature survey
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performed per STPs 1059 and 1061 to study the bulk area i
temperatures and temperatures in the vicinity of control room i
panels.
The tests showed moderate variations of 3-4 C, At the l
l time of this inspection, the licensee was evaluating the data
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to determine the need for further action.
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(3) LER 86-13-LO (Closed) - Actuation of CR HVAC Due to Spurious i
High Temperature, Spurious Radiation Monitor Spikes, and Dead Bus Transfers
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The report detailed additional problems with spurious automatic actuations of the system as indicated above during June and July 1986.
As indicated above, the high temperature problem is being addressed.
The radiation monitor spikes are bs?ievea due to electromagnetic induced pulses in the detector circuitry and are addressed by Problem No. 26.0154.
The licensee has
attempted to correct the problem by improved grounding of the j
circuits and use of copper shielded detector enclosures; the j
effectiveness of the f atter modifications has not yet been l
determined.
Actuation of the system during " demi bus" realignment of electrichl systems is beir.g followed as Problem No. 28.0483.
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The licensee s plans include attempted replicatica of
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conditions during integrated loss of power testing.
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(4) LER 86-23 (Closed) - Control Room HVAC Could Not Perform Curing Low Offsite Grid Voltage Condition l
The CR HVAC system is equipped with flow control urjts having control power supply units.
During engineering reviews associated with ongoing system modifications, the licensee determined that the existing power supply units would not maintain adequate flow control unit voltage under design basis
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degraded grid voltage conditions.
i The licensee has initiated ECN-0911, tracked as SRTP Problem No. 26.0153 to replace the power supp1v units prior to plant restart.
Further, the licensee committed to complete a~revtew of all loads utilizing 120 VAC Class 1 power supplies to ensure adequate voltage availability prior to restert.
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F.
LERs Not Ready for Inspection r
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The following LERs were selected for inspection during the current period but could not be inspected due to incomplete licensee actions as detailed below.
The licensee's program for disposition of NRC findings and report commitments involved completion of disposition actions by the responsible licensee "line" department, forwarding documentation of l
disposition to the SMUD licensing group and verification of action l
completion by the QA department.
As noted below, some of the items
l had not yet completed this process and the licensing gr6up was I
awaiting notification of completion from the line departments.
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In the other cases (as noted), the licensee's committsd action dates or followup report dates had not yet been reached.
(1) LER 86-16-L1 (0 pen) - Decay Heat System Trip Following an Arc j
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in a Sump Level Indicator" '
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This report revision, issued May 1, 1987, documented status to
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date and advised the NRC that diagnostic testing intended to
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reenact the event under controlled conditions had been postponed until more suitable plant conditions exist.
The report advised that the testing is expected to be completed and a supplemental report submitted providing the results
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during July, 1987.
The licensee had not yet collected l
documentation of nor verified interim actions taken to date.
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(2) LER 86-17-LO (0 pen) - Diesel Generator Automatic Loading During LBLOCA Did Not Account for AFW Pump Runout The report addressed a design error resulting in the potential overload of Emergency Diesel Generators by automatically
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started Auxiliary Feedwater Pumps, t
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The licensee advised that the corrective and preventive actions were complete for this item except for publication of the final
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l report of design reviews, collection of documentation, and QA l
verification of implementation.
(3) LER 87-06-L1 (0 pen) - Motor Operated Valves (MOV) Operability Problems The report documented the states of a program develope-d in response to IE Bulletin 85-03 and various problems identified during licensee valve inspections and maintenance resulting from the bulletin.
The above m v 41on of the LER committed to provide a supplemental report by July 15c 1987, documenting completion of ongoing valve testing; a final report will be provided by September IS, 1987, to update the final test status and the i
licensee's assessment of safety consequences during power
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operatione for the "as found" configurations of the valves.
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(4)
LER 87-12-L1 (0 pen)
,_F ilure to Perform Surveillance
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Procedures by Technical g edification Pequired Da g
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Revision 1 to the report, issued on May 12, 1987, revised a
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previous li(.ensee commitment for completion of the root cause I
investigation for the event.
The licensee rescheduled completion of the root cause investigation and prescription of
s corrective acticns via a supplemental report free May 15, 1987,
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to July 30, 1T7.
. i No violations were identified.
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3.
Generic Letters t
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Generic Letter (GL) 86-05 (0 pen) - Implementation of TMI Action Item i
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II.K.3.5, Automatic Trip of Reactor Coo 1 ant Pumps
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GL 86-05 requested plant specific information of licensees who f
adopted the Babcock and Wilcox Owner's tiroup (BWOG) methodology and i
generic analyses for the subject TMI Actica 1trn.
The licensee
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submitted its response to GL 86-05 to NRC:NR2 un April 28, 1987 (letter JEW 87-491); at the' time of this inspection, NRR's review was in progress.
This inspection was not directed at confirming the technical adequacy of the analyses, methoda or technical positions taken by the licensee in its respanse.
It was directed at performing field confirmation of the inforination jargvided in the response including:
Conformance of licensee procedures with the esthodblogy presented in the BWOG keport No. 77-11490G1-00, " Analytical Justification for the Treatment of RC Pumps Following Accident Conditions _, February 1984."
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Conduct'of personnel training on the revised procedures and accident diagnosis methodologies.
Conformance cf-the plailt specific data provided in the response with plant rpecific as-built conditions.
Implementation of modifications :liscussed in the response, including installation, testing, and procedure updatss.
Inspections and reviews were conducted in the following subject-creas corresponding to the subjects of GL 86-05 and the response:
(1) Determination of RCP Trin Criteria j
RCP trip criteria are based upon subcooling margin.
The response provided instrumentation uncertainties associated with the Safety Parameter Display System (SPDS) and Subcooling
Margin Monitor (SMM) which ert the primary iratruments used to I
determine subcooling margin.
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The response stated that the instrument uncertainties under
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I adverse, post accident containinent environntsntal conditions
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were as follows:
SPDS Pressure Uncertainty 301 psig
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SMM Pressure Uncertainty 1270 psig j
The resp;ose further noted that the values were based on the l
manufacturer's specification of accuracy uitfer post accidest I
conditions.
The lic9nsee further stipulfsted that the values were very conservative in that the operator determination to trip RCPs under accident conditions would occur early in the
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accident scenario before adven e design basis environmental
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conditions would affect the inst uments.
l Tne inspector was advised that this data (and corre ponding
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temneratur'a instrument uncertainties) was~i gluded in the licensee's response in a preliminary form from 'a calculcthn which had not been fully approved et the time of response issante.
Subsoquencly, minor errors were found in 'che calculations which have been corrected and are again in the
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Feview and approval cycle.
The new SPDS and SMM preliminary l
values are 300 psig and 281'psig,'respectively.
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The inspector reviewed tne unapproved original and corrected calculations confirming that they used plant specific instrument data and appeared reasonable.
The inspector not' d e
that the calculations conservatively used 5.quare Mot of the sum of the squares analysis methods and included maximum design dnta uncertainties tior.each component of-the respective instrument strings.
The inspector advised the licensee that corrected data must be pubmitted to NRC, revising their osiginal respanse upon final calcu% tion approval.
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i The inspector also reviewed Emergency' Procedures E.01 through
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E.07 finding that the procedures included operating curves
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l-displaying various pressure-temperature operating regimes.
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These curves were based on B&W Document No. B&W 74-1122501 and graphically displayed the' offset curve of " variable'subcooling.
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margin" (V$M).
The licensee's' submittal indicated that the VSM
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y offset corresponded to an equivalent offset in the SPDS display and included margin due to instrument uncertainties.
That.is, the VSM curve represents the margin necessary to be maintained
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above saturation conditions to' ensure that the~RCS remains I
i subcooled when c0nsidering instrument error.
The graphs showed i
a required pressure margin of'about 100 psig at RCS i
teciperatures.of 200 F and ot about 200 psig at 600 F.
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The inspector '=ted that the variable values on the B&W i
generated curw did not agree with the licensee's plant'.
j specific instrument uncertaintin.
Further, the licensee's J
response indicated that the instrument uncertainties were time dependent as the containment environment degraded during the accident.
The Erhargency Procedure curves did not appear to contain any treatment uf this time dapendency but were also applic.able for protracted accident scenarios which might result-in protracted operation ~under adverse containment environmental conditions.
Whfn1 requested to reconcile and explain the differences between.
the B&W curve and the licensee'.s calculation, the cognizant
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engineer advised that B&W.had prepared an equivalent plant
l specific calculation at the licensee's request'but, due to a-
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l contractual dispute, the vendor has not provided the calculation to the licensee.. This engineer had briefly reviewed; an informal copy of these calculations, believed that they did not favorably agree with the' equivalent SMUD calculations, and did not know whether the B&W calculations were (or would.be)
used to reconcile the plant specific questions regarding the
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VSM curve of these B&W calculations.
The inspector requested the SMUD Licensing Group to provide j
additional information which would confirm that the Emergency Procedure curves indeed represented plant specific conditions and accommodated the time sensitivities of> degraded containment environmental conditions.
On June 24, 1986, the Licensing Group issued Memo NL 87-881 documenting the information needs;
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for licensee followup.
This portion of the review of GL 86-05 will remain open pending
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availability of the additional information requested of the licensee.
(2) Oct.ential Reactor Coolant Pump Problems
The licensee's response discussed operation of the RCP Component Cooling Water (CCW) supply and RCP Seal Injection and
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Return services.
The inspector again reviewed the above
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Emergency Procedures and Plant operating Drocedure A.2, Reactor Coolant Pump System, confirming the accuracy of the assertions made in the licensee's. response.
The response noted that air operated valves in the CCW lines to the RCPs will be equipped with two backup sources of operating air curing the current outage. The inspector reviewed FCN
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R059, " Backup Air to MSS, IAS, FWS, and CCW Valves" including
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the Facility Change Safety Analyses, Design Basis Report, and
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relevant drawings.
The inspector also reviewed relevant
portions of STP.774, the preoperational test of toe backup air
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supply.
No discrepancies were' identified.
The response further described the components required to trip the RCPs.
The inspector reviewed portions of the licensee's response involving control room and selected switchgcar locations by comparing the response descriptions to the as-installed controls and equipment.
No discrepancies with respect to the licensee's response were identified.
(3) Operator Training and Procedures The inspector confirmed that the Emergency Procedures had been revised as asserted in the response and that training had been provided the operators.
The inspector reviewed two example Simulator Training Scenarios
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(Nos. 24K7200 and 2K6800) involving loss of feedwater and LOCA operations confirming that they included specific training on l
the RCP trip criteria, diagnostic methods, and procedural
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t requirements.
The scenarios were included in the last simulator training perioo (February 9 - March 20, 1987) and the next scheduled period (June - August 1987).
All licensed operators attended the prior session.
The inspector further confirmed that the pertinent subject
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matter was included in the " restart tt aining" program, e.g.,
ler, sons OD24D4900, E0P Technical Basis and OD24D400, E0P Philosophy and U:c.
ine licensee advised that these lessons and tiie simulator scenarios above were only several of many l
examples in which relevant training is provided, i
Except as noted in A. above, the inspector had no further questions and no violations were identified The inspector advised the licensee that the acceptability of their response also remained subject to completion of the NRC:NRR review.
No violations or deviations were identified.
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I 4.
10 CFR 21 Reportis A.
10 CFR 21 Report 87-12-P (Closed):- TDI EDG Static Exciter Voltece
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Regulator (SEVR) Reset Circuit
.a On March 11, 1987,- IMO Delaval' reported'that the SEVR system is-f designed to deenergize upon receipt of a normal shutdown signal and i
to reset when the engine speed falls below 200 rpe.
If an emergency j
start signal is' received 'oefore the engine coasts from 450 rpn,
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(synchronous speed) to 200 rpm, the engine will accelerate to normal F
operating spead, but the' exciter would.not reenergize, preventing
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generator output.
The 1(censee assigned responsibility for disposition of the report to the Electrical Engineering Group via an Action Item Memo and the.
item was being tracked as TDI Problem Report File No. 55 as a.
?l preoperational deficiency.
The inspector discussed the report and licensee plans with the cognizant licensee engineer who advised that the licensee is currently evaluating a vendor proposal for a corrective ' design
change, including comparison ~ to crrrective actions successfully
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taken at another' facility.
The engineer and inspector reviewed the circuitry involved and preliminary modification plans..
The inspector further confirmed that'the GM/8ruce diesel generators also installed at Rancho Seco.have been evaluated for similar
defects.
This evaluation was performed pursuant to IE Information Notice 86-73 applicable to the GM diesels and as documented in Mento GVC-87-112, ' dated January 21, 1987.
l The inspector and engineer also discussed post modification testing considerations for the SEVR with the engineer acknowledging that'the
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modification will require evaluation with respect to currently scheduled SRTP testing of the EDGs per Regulatory Guide 1.108.
The l
engineer stated that the modifications and' testing would be coordinated to ensure that SRTP testing would not be invalidated by.
the planned modifications.
The licensee appears to have acceptebly integrated the disposition i
of this report into the applicable management control systems.
This item is considered closed.
No violations or deviations were identified.
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10 CFR 21 Report 87-11-P (Closed) - Discrepancy Between As-Tested f
Net Positive Suction Head (NPSH) Curve and NPSH Curve Provided by-Vender for Spare Decay Heat Pump
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In February - April 1987, the licensee identified a suspected casing leak on the "A" Decay Heat Pump.
Although the suspected leak was later 'found to be condensation / spray from an adjacent pump drain
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piping leak impinging on the pump casing, the licensee purchased a-replacement pump from the cancelled Midland Nuclear Station.
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The report documented a discrepancy between the pump's original performance test curves provided by the pump manufacturer and the
results of SMUD contracted testing on March 4, 1987.
The latter
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testing results agreed with the manufacturer's curves with the exception of NPSH versus capacity.
The results indicated that the pump's NPSH requirements exceeded that available had it been j
installed in the Rancho.Seco Decay Heat System.
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The pump manufacturer advised that the ancmalier, are apparently the
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result of excessive internal pump clearances.
The report further I
indicated that the pumo will not be placed in service until
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modifications are performed to correct the deficiencies.
The inspector discussed the report with the cognizant licensee
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engineer who advised that-there are currently no plans to modify the
pump since the existing "A" pump did not actually have a casing
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leak.
The spare pump is being held by the manufacturer pending shipping instructions from the licensee.
Apparently the spare pump j
was also found to have a higher driver horsepower requirement than
the existing pump motor, further complicating its use.
The engineer j
stated that the. pump would bc warehoused until &n alternate, likely
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non-nuclear, application could be found for it.
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The engineer further advised that the discrepancy may not have been j
identified by normal replacement testing per ASME Section XI had the J
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shop testing not been performed.
The inspector further verified l
that the licensee plans to obtain new performance data for the existing pumps.
Tests STP.1033A and B include provisions for
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verification that, at miniinum flow operation, subcooled conditions l
are maintained at the pump suction and that sufficient NPSH exists.
The inspector also discussed receipt inspection requirements with the licensee QC supervisor and reviewed AP.605, General Warehousing, Revision 12. confirming that adequate controls were in place to ensure that the discrepant pump would be quarantined upon eventual receipt and not released for safety-related service without disposition of the subject conditions.
This report is closed.
No violations were identified.
5.
Previous NRC Inspection Findings A.
(0 pen) (Unresolved) 85-31-01, - Procedures for USAR Revisions In Inspection Report 85-31, the inspector identified that in l
Amendment #2 to the Updated Safety Analysis Report (USAR), the licensee inadvertently omitted revisions to section 6.1 related to modifications made to the HPI system, although the changes were l
included in section 9 of Amendment #2.
The licensee committed to I
review and identify their plans for changing their procedures to improve the accuracy and timeliness of reports of facility modifications.
The licensee's corrective action plans were reviewed again during Inspection 87-11, finding that they included plans for a USAR revision procedure to be used for all revisions after 1987 i
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and which would include provisions for a full review of the USAR on i
a rotating basis.
Licensing Department Administrative Procedure, LDAP-003, USAR Revision Control, Revision 0, was issued on June 8, 1987, and reviewed by the' inspector with regard to the above item.
The procedure provides extensive guidelines for capturing changes to the plant and procedures for USAR incorporate 0n and appears responsive to prior licensee problems.
However, no guidance is provided to ensure that all multiple USAR sections potentially affected by a single design cr proc'edure change
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are updated nor does it address the rotating review of the USAR.
i The inspector met with the SMUD Licensing Supervisor on June 24,
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1987, ar.d was advised that the procedure was issued in its present form to previde improved guidance for ongoing 1987 USAR re/ision but that the licensee considered it only partially complete.
The licensee is currently refining the detailed plans for the above oraissions and other, unrelated improvements to be addressed in a future revision.
This item will remain open, pending completion of the above licensee actions and NRC review of the eventually revised procedure.
B.
RV - MA-4 (Closed)
l SMUD - 16c(4)
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EDO - 4a l
l Verify Operability of Manual Valves and Remote Operated Valves.
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Perform Inspections to Assure IntodrIty of Packing an'd Yarify. Proper
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Assembly of Manual Operators Incl,uding Setting of "Neutrs1 Position-T and Mounting Devices Inspection Repot ts 50-312/86-07, 87-11, and 87-16 documented NRC inspection of the licensee's inspection, repair and testing of manual and remote operated valves.
Inspector concerns regarding manual valve inspection and preventive maintenance program were identified in Inspections Reports 87-11 and 87-16:
(1) The listing of 142 valves selected by the licensee as critical to plant operations appeared to have been developed informally rather than via a formal, pre planned review process.
The suitability and completeness of this listing has been referred to NRR for further evaluation.
(2) The inspections of the valves had been performed by maintenance personnel with little or no QC involvement.
Many valves were accepted as-is with only minor upkeep performed.
No verification of individual valve conditions was available.
(3) AP 650, PM Program, categorized manual valves in QA Class 1 systems that are required for controlled safe shutdown as Categories 1 and 2A.
Preventive and corrective maintenance had l
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been performed on or planned for the Category 1 valves but the licensee's program did not include requirements' to inspect and
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maintain the Category 2A valves prior to restart.
(4). Although the licenses had reported completed inspections and (
corrective maintenance on the 142 " critical" valves of (1)
above, sampling reinspection and observation of valve operation-by the NRC inspectors found significant valve condition discrepancies including fastener thread engagement
deficiencies, valve handwheel marking and position indicator
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discrepancies, and packing and mechanical joint leaks.
As immediate corrective action, the Maintenance Manager and QC Supervisor initiated an immediate 100 percent reinspection of the valves.
The licensee actions taken since Inspection 87-16 were reviewed during this inspection.
fhe licensee's reinspection program of (4) above included major l
revision cf inspection data sheets to include characteristics such
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I as discovered during Inspection 87-16.
All-inspections were
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l performed by QC inspectors.
Deficiencies were noted on the data
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I sheets and Work Requests were issued for entrective actions.
Additionally, the licensee continued with 100 percent inspection of i
the Category 1 valves and had instituted a 100 percent inspection of 106 Limitorque manual geared operators subject to handwheel / position indicator discrepancies found in Inspection 87-16.
These inspections and resultant corrective maintenance were continuing at j
the time of this NRC inspection.
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The inspector reviewed the inspection data sheets for all of the 142 " critical" valves reinspected by the licensee.
Sixteen valves were selected for visual inspection by the inspector, six of which were equipped with manual geared operators.
The latter valves were exercised in the presence of the inspector to verify position indicator operability.
The inspector confirmed that the licensee
had initiated corrective action for all valve discrepancies.
The Category 1 valves not included in the 142 " critical valves" list are being progressively inspected with the task being centrolled via
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Work Requests and similar inspection data sheets.
The inspector also selected a sample of Category 1 valves confirming that their inspections had been initiated or completed and corrective actions initiated for discrepancies (6 Work Request Packages involving 82 of 219 Category 1 valves).
Based on the sample reviewed, the licensee's inspection process l
appears to be effective in identifying and correcting valve
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deficiencies.
However, the inspector noted that the licensee's completion status keeping was cumbersome and required manual searches of many Work Requests each applicable to sometimes dozens l
of individual valves.
Although all inspector questions were
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acceptably resolved by the licensee, the inspector observed that the licensee will need to exercise care to ensure that all valves
are addressed as committed to the NRC.
The above mentioned NRC concern that Category 2A valves were not planned to be inspected prior to restart, as specified by AP 650, had been addressed by the licensee via the issuance of a new i
Maintenance Administrative Procedure, MAP-0009, Preventive j
Maintenance Program, on June 24, 1987.
The procedure was in the i
process of distribution for initial implementation at the close of j
this inspection.
This procedure redefines the valve categorization such that all valves previously in Category 2A per ape 650 which were also required l
for safe controlled shutdown, maintenance of the plant in cold j
shutdown, or to mitigate the consequences of a radiological release,
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etc., would be included in Category 1.
Category 2A would include j
the remaining valves in QA Class 1 systems (not included in Category
1) which could have an impact on system process flows and which are
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>2-1/2" in size.
The procedure further makes preventive maintenance
_of Category 1 items mandatory.
The licensee advised that the AP-650 valve category lists would be j
revised to re-categorize the valves accordingly.
The schedule and
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personnel resources for this activity were being arranged during this inspection.
Notwithstanding the above, the current licens,ec program appears to address NRC concerns regarding pre-restart upkeep j
of critical valves.
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This item is closed; no violations or deviations were identified.
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l C.
(Closed) (Inspector Follow-up) Item 87-06-10 - Adequar.y of QC l
Inspector Training
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During followup of Allegation RV-86-A-0099, the inspector found_that site specific QA/QC training for QC inspectors consisted of individual reading of QA/QC procedures and a written exam.
No formalized classroom training was given nor were inspectors given
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adequate opportunity to obtain insight to site specific
requirements.
Since that inspection, the licensee has initiated action to provide additional training to permanent (SMUD employee)
inspectors as further discussed below.
The basic requirements for QC inspector training and certification were provided by QAIP-18, Quality Control Inspector Training and Qualification Requirements.
This procedure described a combination of training (primarily self-study as implemented), on the job training, and certification testing which met the licensee's regulatory commitments to ANSI N45.2.6-1978.
Although the licensee intended that this procedure remain the primary training instrument for non-employee (contractor) inspectors, its provisions are being supplemented by two other programs.
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j First, staff QC inspectors have been included in the Technical Staff l
Training Program (which is part of the INPO accredited site program)
as implemented by Program Description TS21Z0000.
This program
included formal classroom instruction and student performance
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measurement in the subject areas of engineering fundamentals, instrument and controls fundamental, plant systems and compcnents, plant operations, plant physical and procedure changes, codes and I
standards, regulatory changes, and industry lessons learned.
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The inspector reviewed specific training session examples including training on NCRs'(administered to about 360 personnel), training en procedure changes per RSAP-1201, and the scope and schedule of the next cycle of general technical staff training.
In addition to the " generic" staff training, specific staff
assignment Training Check Sheets (No. TS-21H-0500) had beer, j
developed tn include QC specific on-the-job trainittg activities with
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corcpletion accountable to the plant training department.
Each staff inspector's experience and prior training would be
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evaluated by the training department.
Individual exemptions may
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result frem specific classroom or on-the-job training requirements
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from the evaluation.
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QC participation in the Technical Staff Training Program had begun J
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only recently.
The QC training resource material above was available and two (of twenty employee) inspectors began training on June 22.
The remaining 6taff were to be enrollec; in subsequent training cycles.
l The second effort underway was in the planning and scoping stage.
i The licensee was participating ir. a regional _ group of utility QC organizations seeking to develop an inspector training program based upon the INPO " Principles of Training System Development" (INP0 85-006).
The licensee noted that INPO has not established specific accreditation standards for QC inspector training and that the program currently under develuptent exceeded the current regulatory requirements and INPO guidelines.
Eventually, this program was intended to provide equivalent generic training to each of the groups' nuclear stations, permitting reciprocal acceptance of training and certifications for both contractor and staff inspectors.
The licensee has begun development of position specific task lists,
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surveys, and program outlines for the above program.
Concurrently, i
training material from the other utility group participants was being evaluated.
'No target date was available for full implementation, partially because of the need to evaluate existing material acceptability prior to making implementation and schedule decisions.
Completion of the SMUD specific program description was,
however, scheduled as WorkTrak Items 870674 and 75 for the last quarter of CY 1987.
The licensee does not intend to provide enhanced training to i
contract inspectors, The licensee currently employs fifty seven
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contract inspectors.
This number is declining as major outage work
is completed (down from about 100 inspectors carlier in the outage).
j Memo WCK 87-024, W. Koepke to S. Knight, dated May 28, 1987, documents the QC Department's position that contract inspes: tors will
meet the requirements of QAIPc18 above and are subject to close I
direction by lead SMUD inspectors.
This training incitdes initial certification and refresher / change training as required.
No changes
are planned in tnis program.
I Based on the foregoing, the licensee's plans and programs for permanent employee inspectors appear adequate.
The licensee's training for contract QC inspectors appeared to meet the applicable regulatory requirements and commitinents, and is therefore also j.
adequate, but it was substantially less extensive than that afforded (
permanent inspectors.
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This item is closed.
No violations or deviations were identified.
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(Closed) (Inspector Followup) Item 86-17-03 "Determing Applicability
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of 10 CFR 21 Reporting Requirements to Re'sults of Electrical i
Termination Inspection Program.
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The licensee initiated an electrical termination inspection orogram as a result of the December 26, 1985 trip in which a faulty compression lug termination caused the loss of the Integrated
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Control System (ICS) control power.
The program was initiated to j
provide assurance that potential lu0 failures were identified and i
corrected, j
i The licensee's activities were previously reviewed during Inspections 86-17 and 87-08.
During those' inspections, the status of the licensee's program was reviewed,' including the types of
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deficiencies identified.
On June 17, 1987, the licensee's final l
program report was issued detailing the program results.
As part of
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this activity, the licensee concludeel that the deficiencies found did not constitute defects in basic components as defined by 10 CFR 21.
On July 23, 1986, the licensee submitted a Licensee Event Report identifying the potentially generic 19plications of the program findings.
In this inspection, the inspector reviewed the pertinent portions of the June 17, 1987, report and met with the licensee's Electrical Engineering Group management and QA personnel to discuss the program results.
The inspector noted that no actual termination functional failures were found.
Although the licensee's report characterizes numerous inspection and test findings as " failures", it is'in the context of the item having failed tne inspection criteria, not that the 1 tem failed in service or represented an imminent failure.
The licensee was able to demonstrate, to the inspector's. satisfaction, that no report pursuant to 10 CFR 21 was required.
This item is, therefore, closed.
No violations or deviations were identified.
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(Closed) (Inspector Followup) Item 87-08-01 Licensee Disposition of'
" Trimmed Electrical Lugs" Found During Electrical Termination -
Inspections During the licensee electrical inspection program discussed in Item 86-17-03 (above), some termination lugs were found to be mismatched with the terminal boards such that the lugs had.been trimmed to fit i
them in the boards. These discrepancies involved vital power cable terminations.
Tne inspector. reviewed the licensee's disposition of this item as documented in.the June 17, 1987,. final report and via discussions with licensee engineering management..The licensee had added the inspection requirements for this item to Quality Assurance Inspection Procedure.(QAIP) 15, Electrical System Upgrade Program,
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the controlling document for the overall -inspection effort.
The final report documents the replacement of 31 terminations and disposition of 14 Nonconformance Reports and 14 Engineering Change Notices to corre::t identified deficiencies for inspections in the ccatrol/ computer ro?m.
The licensee's plans for a continuing program to include.the inspection of vital power cables outside the control / computer room identified as having a high probability of field modified (trimmed)
lugs.
The licensee has issued 80 Work Requests to inspect these a*eas with about half of'the work complete.
An additional 80 work requests hcve been issued to inspect shielded cable terminations; five remain to be completed.
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The licensee's programs appeared adequate to assure a con' trolled-l completion of reasonable inspections for the identified. defects.
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This item is considered closed.
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l No violations or deviations were identified.
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F.
(0 pen) (Inspector Followup) Item 86-37-03 (Partial) Adequacy of Location for CR HVAC Radiation Monitor Detectors R15701 and R15702.
During Inspection 86-37, the inspector noted that the subject l
monitor detectors are not installed in the CR HVAC ductwork but were l
installed on a wall approximately ten to eighteen feet from the HVAC l
air handler intakes.
Each'of the channels (R15701 and R15702)
actuate an individual train of the essential HVAC system.
Although the licensee's staff considered the locations to be i
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The inspector noted that the j
instaliatir-Joes not appear to be consistent with.the general h
principles o, HVAC system design discussed in " Industrial Ventilation," 18th Edition, 1984.
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During this inspection, the licens'ee provided the inspector an Office Memorandum"(D. Marsh to D. Cox, "CR/TSC Normal HVAC Intake Rad Monitors R15701 and R15702," dated May 22, 1987), documenting,a qualitative evaluation of the. detector locations.
The memo discussed the calibration geometry, nuclide sensitivities, nuclide beta radiation energies and ranges, source terms, and detector location.
The memo's general conclusions were that:
(1) The detectors are ' calibrated for ~ Kr-85 and Xe-133 beta radiation which'is of relatively low energy and range.
(2) The other beta emitters expected in the post accident noble gas mixture are of significantly higher beta energies and ranges and will be present in significant concentrations.
(3) Although the detectors are calibrated for Kr-85 and Xe-133, they will "see" the nuclides of (2) in.that their beta. ranges
in air are between 30 and 50 feet (approximately.the full width of the building air intake louver wall.
(4) Based on the above, the memo summarized that the " detectors appear to be mounted in an acceptable configuration," that the
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calculated setpoint is very conservative, but the upper, monitor j
"should probably be lowered at some. time to' optimize'its j
location relative to. air. in-flow pathways."
The inspector reviewed the above memo, various system descriptive information, and Setpoint Calculation Z-ZZZ-M1751,' dated October
.J 1985, and discussed the matter with licensee engineers responsible
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for development of_the memo, the calculation,'and system j
configuration' control.
The above calculation determines the actuation setpoint'for isolation of normal CR HVAC and actuation of essential CR HVAC.
As indicated, it is based on the detectors' sensitivities to Kr-85 and Xe-133 expressed as counts per minute per microcuries per cubic
centimeter.
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The calculation derives equivalent isotopic concentrations at which
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the equipment should actuate based on desired dose reduction factors and detector characteristics.
During this review, the inspector noted that the. isotopic-
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sensitivities appeared reversed in the equations (i.e.,'the higher i
energy beta emitter, Kr-85', resulted in a smaller detector response
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than the Xe-133 (about' one-half the maximum beta energy'of; Kr-85)..
l When questioned, the cognizant licensee engineers acknowledged that they had discovered the same error during a previous peer review of the calculation, identified it for correction, but had not yet
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t processed a corrected calculation.
The licensee stated that a'
i formalized. recalculation would be performed.
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I The engineers further advised that informal, bounding calculations had established that the error resulted in negligible impact on the actual setpoint.
The inspector confirmed a minor error effect by independent calculation.
The licensee's identification and tracking of the discrepant calculation appeared to be informal, relying on oral communications.
The inspector noted that this error and informal control of design processes was similar in nature to findings identified during.the " Augmented System Review and Test Program Inspection" documented in NRC Inspection Report 50-312/86-41.
With regard to detector location, the inspector confirmed the accuracy of the assumptions and isotopic data provided in the May 22 memo using the " Radiation Health Handbook," U. S. Public Health Service, revised January 1970.
The inspector further consulted with the cognizant Region V radiation protection specialist.
The licensee's evaluation did not, however, address air flow considerations, beta energy attenuation vs. range vs. detector sensitivity, or the specific geometry in quantitative analyses.
The licensee memo itself appears to support the need for a more deterministic analysis to ensure that the detectors will "see" and react to radiogas levels as necessary to limit control room personnel exposures.
Subsequent to the inspection, on June 30, 1987, the licensee, the i
inspector, and a representative of the Region V Radiation Protection Section conducted a telephone conference call (at the licensee's request) to discuss the inspectors' concerns and licensee plans for addressing them.
The information contained herein was reiterated and the need for a quantitative analysis of the detector locations was emphasized.
The licensee agreed to perform a quantitative analysis using appropriate input data.
I No violations or deviations were identified.
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(0 pen) Violation 86-07-01 Civil Penalty Violations Resulting from Inspection of the December 26, 1985 Event i
This item collectively addresses eleven violations identified during l
Inspections 50-312/86-06 and -07 and an open inspection followup item of the same tracking number.
The violations are summarized as follows:
l (1) RCS cooldown rates violated Technical Specifications.
(2) Failure to correct identified design deficiencies leading to loss of Integrated Control System Power and to develop procedures which would have lessened the severity of the ensuing transient.
(3) No written procedures existed for securing high pressure injection following safety features actuation or for manual
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4 emergency operation of an auxiliary feedwater system control valve.
(4) No written procedures existed requiring periodic maintenance of an auxiliary feedwater system valve.
(5) Emergency Operating Procedure E.05, Excessive Heat Transfer, was inadequately implemented during the event.
(6) Alarm procedures for evacuation of personnel for high gaseous activity monitor alarms were inadequately implemented.
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(7) Procedures for evaluation of unplanned radioactive releases
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were not implemented.
(8) Emergency Plan in plant notification procedures were inadequately implemented.
J (9) Emergency Plan procedures for notification of state and local governments were inadequately implemented.-(2 examples)
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(10) Emergency Plan Procedures for implementation of radiation j
monitor alarm setpoints were inadequately implemented.
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l (11) Further, Section 2.E of Inspection Report 50-312/86-07 (Item
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86-07-01) noted that the items collectively indicated a serious
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l breakdown in management controls which should establish and
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implement important procedures.
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The status of licensee actions taken in response to the above were
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l reviewed during the inspection as described in the licensee's
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response (dated November 20, 1986) to the Notice of Violation and-j Proposed Imposition of Civil Penalty.
The licensee's response provides numerous commitments for corrective and preventive actions for the specific violations enumerated above.
i Some of these response actions apply to one or more aspects of the violations.
The response letter further referred to the licensee's
" Action Plan for Performance Improvement" separately submitted to
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NRC.
I However, the inspector noted that the response did not include a coordinated response to item (11) above, i.e., although each individual violation had been addressed by the licensee's response, it did not explicitly address the collective breakdowns described.'
Further, the licensee's preventive and corrective action commitments involved many individual action items, each requiring documentation of implementation.
The SMUD Licensing Group is tasked to collect and collate this documentation; the SMUD QA Department is similarly tasked to verify adequate implementation.
At the time of this inspection, some of the actions were still in progress but only a few of the completed actions were documented by
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't the plant staff and forwarded to the_ Licensing Group.
The licensing group was also informally logging and maintaining status on the j
items but only a partial status was available.
The inspectors met with the Licensing Group Supervisor on June 24
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and 25, advising him of the following:
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Because of the seriousness of the issues identified by the a
violations and open item, NRC Region V considered the items to i
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warrant high licensee attention and upper management review of j
i corrective action adequacy.
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The apparent lack of documented completion available in the I
Licensing Group files indicated less than desirable attention
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and response from the responsible plant departments in either
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implementation of the licensee response or its documentation.
The licensee documentation available to date does not address
item (11) as discussed above.
The Licensing Group Supervisor
stated that the licensee considered the response to implicitly Q
address item (11) and requested further clarification of NRC's j
concerns.
l The inspectors emphasized the licensee's need to provide a
comprehensive and coherent representation of how the licensee perceived and formally identified the programmatic weakness identified in item (11) and explicitly how the committed licensee actions address ~the causal factors that resulted in
the program breakdowns.
Specifically, the licensee has not addressed the causes of i
multiple program breakdowns in a manner which demonstrates that the committed actions will be effective in correcting the l
breakdowns across the licensee organization.
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l The licensee representative emphasized the licensee's desire and intentions to responsively address these concerns and stated that the matter would be reviewed by licensee management and suitable actions initiated.
Pending licensee's completion of its corrective action, the violation was left open.
6.
Unresolved Item Unresolved items are matters about which more information is required to
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l determine whether'they are acceptable or may involve violations or deviations.
One new unresolved item identified during this inspection is discussed in paragraph 2.D.
7.
Exit Interview The inspection scope and findinCs were summarized on June 25, 1987, with those persons indicated in paragraph 1 above.
The inspector described
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the areas intpected and discus ed'in detail the inspection fin' dings.
No dissenting coinments were received i'ron the licensee.
The 'ollowing new
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item was, identified during this inspection:
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- Violation?87-21-01-Failure to Perform Cable Rcuting Inspections (paragraph 2.A).
thiresolved Item 87-21-02 - hccuracyofCabieAmpacity' Calculations (paragraph 2. A).
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- Viciation 87-21-01 was identified as an unresolved item during the exit I
interview, upgraded during subsequent in-office review and the licensee.
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was notified of the change on July 21, 1987.
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