ML20206R344
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| ML20206R344 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/06/1987 |
| From: | Ang W, Dangelo A, Andrew Hon, Miller L, Morrill P, Myers C, Perez G, Ramsey C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20206R212 | List: |
| References | |
| 50-312-87-06, 50-312-87-6, NUDOCS 8704220149 | |
| Download: ML20206R344 (43) | |
See also: IR 05000312/1987006
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U. S. NUCLEAR REGULATORY COMISSION
REGION Y
Report No. 50-312/87-06
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Docket No. 50-312 i
License No. DPR-54
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Licensee Sacramento Municipal Utility District
P. 0.' Box 15830
Sacramento, California 95813
Facility Name: Rancho,Seco Unit 1-
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Inspection at: Herald, California (RanchoSecoSite),
Inspection Conducted:
Inspectors: M v S' l
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A. J. D'Angelo, Senior Resident Inspector Date Signed
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CJ.' Myers, Resident Inspector Date Signed
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G. P. Perez, Resident Inspector
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Date Signed
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W. Ang, Regional Inspector
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Date Signed
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C. Ramsey, Regional Inspector Date Signed
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A. Hon, Resident Inspector, San Onofre Date Signed
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P. Morrill, Regional Inspector
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Date Signed
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Approved By: i 8 4 ~#7
L. F. Miller, Chief Date Signed
Reactor Projects Section II
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Summary:
Inspection between January 31 and February 27, 1987 (Report No. 50-312/87-06)
Areas Inspected: This routine inspection by the Resident Inspectors, four
regional inspectors, and two consultants.. involved the areas of operational
safety verification, maintenance' surveillance, procurement, receipt, storage,
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and handling, licensed operator training, non-licensed operator training,
Allegations RV-86-A-99, RV-86-A-22, RV-86-A-78, and followup items. During
this inspection, Inspection Procedures 30702, 30703, 37701, 37702, 37703,
38701, 38702, 41400, 41701, 61726, 62703, 71707, 71710, 90712, 92700, 92701,
92703, and 94703 were used.
Results: Two violations related to the control of repair work, two violations
related to the control of safety related spare parts, two violations related
to 10 CFR 50.72 and 50.73 reporting, and one violation related to
nonconformance dispositioning were identified in the areas inspected.
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DETAILS '
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;1. . Persons Contacted.
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a.; Licensee Personnel
I J. Ward, Deputy General > Manager, Nuclear
K. Perkins, Restart Implementation Manager-
G. Coward, Executive Assistant, Special Projects
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B. Day, Deputy Nuclear Plant Manager
J. McColligan, Assistant to Nuclear Plant Manager
18. Bibb, Deputy Restart Implementation Manager
R. Ashley, Licensing Manager.
D. Army, Nuclear Maintenance Manager ~ J
B. Croley, Nuclear Technical Manager- '
G. Cranston, Nuclear. Engineering Manager <
S. Redeker, Nuclear Operations-Manageri
J. Shetler, Implementation Manager, .
T.~ Tucker,- Nuclear, Operationst Superintendent
- M. Price, Nuclear Mechanical Maintenance: Superintendent
L. Fossom, Deputy Implementation Manager'
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*R. Colombo, Regulatory Compliance Superintendent
J. Field, Nuclear Technical Support Superintendent
M. Hieronimus, Asst.-Operations Superintendent -
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S. Crunk,sIncident Analysis Group Supervisor .
3F. Kellie, Radiation Protection Superintendent
S. Knight, Quality Assurance Manager-.
C. Stephenson,. Senior. Regulatory Compliance Engineer
B. Daniels, ~ Supervisor, Electrical Engineering
J. 'Irwin, Supervisor, ~ I&C Maintenance-
C. Linkhart, Electrical Maintenance Superintendent
*R. Cherba, Quality Engineering Supervisor
*T. Shewski, Quality Engineer
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P. Turner, Training Manager
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F. Gowers, Deputy Training Manager
M. Herell, Operations Training Supervisor
R. Considine, Superintendent of Training Support
T. Hunter, Operations Training Supervisor
J. Chwastyk Deputy Operations Training Supervisor
D. McIntire, Restart Training Coordinator
K. Malloy, Non-licensed Operator Program Coordinator
J. McCann, Instructor
H. Heilmetre, Manager, Nuclear Training Services, PSI
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J. Carson, Simulator Instructor, PSI
R. Simmons Senior Instructor, PSI-
G. Wallace, Simulator Program Coordinator
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A. Mcrady, Training Library & Records Supervisor
P. Gibbel, TIMS Expert
M. Turner, Training Records
R. Ashley, Licensing Manager
*J. Robertson, Licensing
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- Other licensee employees contacted included technicians, operators,
mechanics, security and office personnel.
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{ Attended the Exit Meeting on March 12, 1987. -
Management Analysis Company (MAC) Personnel
', 2. Operational Safety Verification I - .- '
Thi inspectors reviewed control. room ope' rations which included access
control, staffing, ' observation of decay heat removal system alignment.
.and review of control room logs. Discussions with the shift supervisors
and operators indicated understanding by these personnel of the reasons
for annunciator indications, abnormal plant conditions and maintenance
work in progress. The inspectors also verified, by observation of valve
and switch position indications, that emergency systems were properly
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aligned for the cold shutdown condition of the facility.
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Tours of the auxiliary, reactor, and turbine buildings, including
exterior areas, were made to assess equipment conditions and plant
conditions. Also the tours were made to assess the effectiveness of
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radiological controls and adherence to regulatory requirements. The
inspectors also observed plant housekeeping and cleanliness, looked for '
potential fire and safety hazards, and observed security and safeguards
practices.
The inspector reviewed Special Test Procedure STP 1011, Decay Heat
Removal Test, for compliance with the Technical Specifications and
potential safety questions which could arise during the conduct of the
test. The licensee addressed a potential concern of boron dilutions
within the Reactor Coolant System (RCS) during the conduct of-the test by
- appropriate valve lineups and sampling. i
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No violations or deviations were identified.
3. Monthly Surveillance
, During this period only. testing related to the Systems Review and Test
Program was inspected.
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The following items were considered during this review: Testing- :
witressed was in accordance with adequate procedures; test .
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. instrumentation was calibrated; limiting conditions for operation were
met; removal and restoration of the affected components were
accomplished; test results conformed with Technical Specification and
procedure requirements and were reviewed by personnel other than the
individual directing the test; the reactor operator, technician or
engineer performing the test recorded the data and the data were in
agreement with observations made by the inspector; and any deficiencies
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identified during the testing were properly reviewed and resolved by
appropriate management personnel.
! . System Review and Test Program (SRTP)
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The areas of inspection included reviews of newly issued Special Test
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Procedures (STPs).revisedAdministrativeProcedures(APs)andwitnessing
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!tne conduct of STP 1019B-(see'below) performed during the inspection
period.
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SPECIAL TEST PROCEDURES
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STP 1019B - Preliminary Functional Testing of TDI Emergency Diesel
Generator GEB2 Phase II
STP 770 - CCW Containment Isolation Valve RCP Interlock
STP 1032 - Nuclear Service Cooling Water Component flow'
Verification
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STP'1034 - CCW to PCW Heat Exchanger Tube Leakage Detection Test
STP 771 - Pressurizer Spray Valve Interlock Functional Test - Test
Outline Review Only
. STP 1044 - Auxiliary Boiler Interlock and Load Test
STP'1011 - Determination of Decay Heat Load
STP 962 - DHS Pump Interlock Functional Test
ADMINISTRATIVE PROCEDURES
AP.44 - Plant Modifications - ECN Implementation
AP.80 - Training of Test Personnel
AP.81 - Certification and Qualification of Test Personnel
, AP.82 - Conduct of Special' Testing
AP.92 - System Review and-Test Program Description and
Organization
AP.93 - System Status and Investigation Reports
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AP.96 - Test Working Group
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- AP.2.27 - Spec'ial Test Procedure Fomat and Content
Test Outlines for a variety of other STPs were also reviewed. *
During the review of newly issued STPs' the inspector identified that some
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procedures contained inconsistencies in their preparation. In
, particular, STP 770 had several steps written that required the test
personnel to perform an action even though the steps were written in the
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context of observation steps which did not require the operator to
perform an action such as manipulating a switch, i.e.: " Verify the valve
cycles closed and open" when it was actually required to " Cycle the
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valve and verify it closes and opens."; In establishing the acceptance
criteria for the aforementioned step, the criteria was stated as "The -
valve opens and closes", was rather than "the valve closes and opens,"
which was the expected sequence of valve positioning. Similar types of
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inconsistencies were identified by the inspector in.other procedures as.
'well. The licensee agreed.with these observations and promptly clarified
the-procedural intent.. ,
During the auditing of the conduct of STP 10198, the inspector noted that
the test performance was. acceptable. Some typographical errors which
required issuance of Temporary Procedure Changes (TPC) were pointed out
to the Test Director by the. inspector. These did not change the
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technical accuracy of the procedure.
Review of preliminary test-results indicated'some. lack of training of the
Test Directors in the preparation of: test documentation. The licensee
,has subsequently issued AP.82, Conduct of Special. Testing, and on
February 19, 1987, fonnally trained all prospective Test Directors on
test conduct and documentation. This training:should improve the
preparation and acceptability of test documentation.
No violations or deviations were identified.
4. Monthly Maintenance .
.The inspector observed a temporary maintenance repair of a leaking pipe
in the "A" train of the nuclear service raw water (NSRW) system. The
leaking pipe was repaired with a rubber patch affixed with four hose
clamps to reduce the leakage. The inspector observed a deficiency tag
enclosed -in a clear plastic bag filled with water from the leakage. The
inspector:was not shown any Work Requests by the licensee which
documented the installation of the temporary repair patch. However the
licensee had written Work Request No. 110755 to effect permanent repair
of the piping. In discussion with licensee representatives, the
inspector found that no nonconformance report had been written on the.
- original deficiency nor had any abnormal tag been written to control the
temporary modification to limit the leakage. from the defective pipe.
This is an apparent violation of the NCR procedure requirements of QAP-17
and Criterion XVI of 10 CFR 50 Appendix B, which require initiation of an
NCR on a nonconforming component (Violation 87-06-01).
Moreover, the temporary repair was an apparent violation of the Aonormal
Tag requirements of 10 CFR 50, Appendix B Criterion III, and AP.26,
" Abnormal Tag Procedures" in that no tag was initiated, nor was any
design control over the repairs invoked (Violation 87-06-02).
5. . ESF Walkdown
-The inspector walked down portions of the main steam line pressure taps
. installed as part of the EFIC system per ECN-5415C. Tne inspector noted
that two of the four pressure tap instrument tubing lines of each main
. steam line did not have electrical heat tracing installed to protect the,
instrument lines from freezing. The inspector reviewed ECN-5415C which
was still open at the time and found that installation of the electrical
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-heat tracing for the instrument lines was not included within the scope-
of work for the Sub-ECN even though it was specified in the Major-ECN and
included as part of the design basis report (DBR) evaluated for the
safety analysis of the modification. A licensee representative stated
that the electrical heat tracing was not to be installed as a priority
' item prior to restart, but rather would be installed in the fall prior tc
the next cold weather season. However, the inspector noted that the
requirement for installation of the electrical heat tracing had been
marginally annotated by Operations in the Safety Analysis, and the MSRC
had reviewed the annotated document. A licensee representative stated
that current controls over ECN preparation did not incorporate methods to
insure that all elements of the modification referenced in the safety
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analysis were actually incorporated within the scope of work of the
Sub-ECNs. The inspector was concerned that this omission might enable
work elements to be overlooked and not be tracked to implementation. The
deficiency was not identified under an Abnormal Tag or an NCR to ensure
that future installation would be performed.
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The licensee's design control system appeared to not have correctly
translated the design requirements as stated in the major ECN to the
detailed sub-ECN, in this case. This also appeared to be a programatic
weakness. (No violation was issued because this error was discovered by
the inspector before the licensee had completed their review of the work
package which was still open and partially unreviewed by the licensee).
The licensee's progress in this area will be reviewed further in
subsequent inspections.
No violations or deviations were identified.
6. Training
A. Purpose and Scope
This inspection was conducted to evaluate the effectiveness of the
licensed operator training program and to determine the adequacy of
licensee followup actions on training open items.
Prior to the inspection, the previous examination of the licensed
operator training program (Report No. 50-312/84-29, November 14-21,
1984) was reviewed. In addition, Inspection Report No. 50-312/86-07
(datedMay 14,1986) documented findings related to the overcooling
transient of December 26, 1985. This report contained two open
items related to training of operators, 86-07-02 and 86-07-06.
Report 50-312/86-07 and the two open items were reviewed for
follow-up.
B. Background:
Following the December 26, 1985 overcooling event, the NRC conducted
a special inspection from February 21 to April 11, 1986.
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During the week of April 7-11, 1986 the NRC staff interviewed
licensed and non-licensed operations personnel to assess the
effectiveness of training after the overcooling event. Although
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the team concluded that the training had been effective, they
'noted eleven items which indicated operator uncertainty in
specalized areas. These items were subsequently discussed with the
Facility Staff and documented as NRC open items 86-07-02 and
86-07-06.
To evaluate the effectiveness of the operations training and
requalification programs, two additional operating events were
chosen to followup through a review of training records, interviews
with personnel, and examination of the licensee's incorporation of
this material into the training program. The first event occured on
November 16, 1986, when isolation of the spent fuel pool leak chase
drain line resulted in an unplanned offsite release. The second
event. selected occurred on November 21, 1986,.when a pressurizer
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heater bundle was burned out due-to energizing the heaters with the
water level .in the pressurizer below the~ heaters.
C. Inspection of Training
Thainspectionof'requalificafionandtrainingwasconductedusing
the six overlapping' ' steps' listed below:
Examination of the current requalification program
Audit of operator requalification records
Evaluation of classroom and simulator training
Interviews with and observations of shift personnel
Interviews with licensee training, operations, and licensing
personnel
-1. ' Examination of the current requalification program
The inspector examined the " Operator Training Program for NRC
Hot License Candidates", TI-80 Rev.1, "NRC Licensed Operator
Retraining Program", T2-80 Rev.2, and the licensee's records of
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personnel in the requalification program.
In addition to this review, the inspector (1) reviewed NRC and
licensee records to determine the pass rates for NRC
requalifications and initial examinations as well as licensee
requalification examinations, (2) determined the status of INPO
Acreditation for the Rancho Seco Training Programs, and (3)
reviewed the qualification of training department personnel.
The inspector determined the pass rates for requalification and
initial examinations for the last three years.
The results are tabulated below:
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' Year JInitial Tests '
Requal Examination
2 Pass Rate' : Pass' Rate..
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1984 100% i
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t100% 75%.
1985'- 82% N/A 97%
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1986 89%~ 82%- 100%
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The inspector also detemined th'a't INP0 Acreditation .had b'een
obtained in April 1986 for the Hot License, Requalification,
STA, , and Non-licensed Operations; training programs.
The inspector found that the instructors were qualified
consistent with the approved _Requalification-Program and
ANSI 3.1 - 1978. However, the inspector also found that all
-but one of the Operations Training Group (approximately 25 out -3
of 26 personnel) were contractors. The inspector asked the
Training Manager when and how many of these positions would be
converted to full time SMUD employees.
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' The'~ Training Manager stated that he had recently requested nine
full time, positions; three supervisors, .and six instructors. To '
date only one position had been_ approved by SMUD (The Simulator
Supervisor position), but that he. expected that the remaining
positions should be approved in the near future.
The inspector concluded 'that SMUD's implementation of the
Re_ qualification Program appeared to be improving since the
previous inspection of it two years ago. Increased staffing,
previous staff B&W experience, and management involvement were
7 strong points of the' current training programs.
Conversely, the Rancho Seco Operations Training staff consists
of over 95% contractor personnel. It is recomended that a
significant number of these positions be promptly converted to f
full time SMUD employees.
2. - Audit of Requalification Program Records
The requalification and training records were audited to
evaluate the licensee's compliance with the regulatory
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requirements. Eleven records were fully audited and several
other records were spot checked.
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The sample was based on
selecting:from each shift: operators, senior reactor
operators,'no'n-licensed operations personnel, personnel who had
been involved in recent significant operational events, and
personnel who were concomitantly interviewed by the inspector.
Documentation for the following: items were audited:
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Annual Written' examinations 1985 -through 1986
Lecture, attendance records for-1986'
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Required Control Manipulations'1984 thro' ugh 1986
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Performance Evaluations 1985 through 1986 -
~ Training required in identified deficient areas
a - Completion and documenta' tion of'Requ' ired Reading assignments
for 1986 and 1987 Resumption of licensed duties. after
' completion of-accelerated training (when required) Review of
j procedures / changes and-self study . During'the audit the sample was expanded to review 100% of the <
records for completion of required reading assignments for licensed
i personnel:(including training department personnel) and non-licensed
operations personnel
The inspector observed that individual requalification files did'
not contain = references to successful NRC SR0 upgrade examinations,
- which obviate the need for an annual requalification exam. The
licensee personnel pointed out that if a R0 passed an NRC test for
[_ SRO, then, in accordance;with the.Requalification Program, the
person:was not required to'also take--the RO requalification
examination. The inspector ~ agreed and suggested that some
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indication of this situation be retained in the individual '
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requalification files.
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The inspector,also' observed that control manipulation number 8,
" Loss of Instrument Air," of the current, approved Requalification -
, Program had.not been completed during the period 1984 through 1986. '
Initial discussions with licensee personnel disclosed that the B&W
<, .(Power Systems International) simulator did not simulate loss of
instrument air for Rancho Seco.* The inspector examined the previous
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! revisions of the Requalification Program, T2-80 Rev.0&1 and found
that the-loss of instrument air manipulation was not required under
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v the old program. Consequently, this manipulation will not be
o overdue until two years after the~ manipulation became a requirement
(December 17,1987).
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While examining the records of completion of. reading assignments the
inspector found that at least 18% of the required reading
assignments for licensed operators were completed more than 30 days
- after they were due. Since 30 days is allowed for completion of the
l assignments'and return of the completion sheet, the subject 18% , L required over 60 days for completion. The inspector found that at
least 16% of the Technical Specification reading requirements
required more than 60 days for completion, while 28% required over
30 days for completion. Approximatly 13% of the non-licensed
operations personnel' reading assignments were overdue by more than r
30 days (60 days from issuance).
In response to the inspector's observations regarding the number
and tardiness of reading assignments, the Plant and Training
Managers stated that effective immediately, when a reading
. assignment or Technical Specification review was past due by two
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weeks the individuals' supervisor would be informed in writing.
-After the assignment was over four weeks overdue'the Plant
Manager / Operations Manager would be informed in writing, and if this
had no insnediate effect, the person would be removed from licensed
duties.
. Licensee personnel also stated ~that starting with reading
assignment 86-11, both licensed and non-licensed operations
personnel would get the same reading assignments.
In general, the records of the training and requalification programs
appeared to be well documented and properly executed. Tracking and
documentation of training completed and training due over the last
two years has greatly improved. No items of non-compliance were
identified. Regarding the required reading assignments, the number
of overdue assignments during the last year appeared excessive.
The licensee's methods and committments to improve this performance
appear acceptable. Continued licensee management attention should
be directed at this area to ensure timely completion of required
reading assignments.
3. Evaluation of Classroom and Simulator Training
This inspection was to determine how effectively the licensee
Operations and Training Departrients dealt with the three selected
operational events and to follow-up the previously identified
trainingitems(NRCNos. 86-07-02 and 86-07-06). A secondary purpose
was to determine if training received prior to the events could or
should have prevented or mitigated the events. In order to properly
evaluate the training material content for technical adequacy, the
response to operating events, and the effectiveness of the training
the inspector reviewed several areas of plant design and operations.
These areas included the Plant Emergency Operations Procedures,
Casualty Procedures, Instrument AirSystem(IAS),theIntegrated
Control System (ICS), the Turbine Bypass and Atmospheric Dump
Valving (TBV & ADV), Feedwater and Auxiliary Feedwater System (FW &
AFW), Auxiliary Steam system, Reactor Make-up System valves, Fuel
Pool Leak Chase Channel Drain System, Pressurizer Level Indication
l System devices, manual valve operations, and the control room panel
modifications. This activity was completed by conducting system
walkthroughs with operations management, review of procedures and
temporary changes, review of Design Change Notices, and discussions
with training and operations personnel.
The inspector audited the following classroom training to verify
that the technical content and instructor presentation of the
information was adequate
- .OD24 C0716 "ICS Modifications Part II
OD24 D2100 "EOP Modifications, Cycle 7 Restart" " Radiation
Protection Refresher
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' Theinspectoralsoreviewddthe'following':coursematerialfor
adequacy in. scope'and technical content. '
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<0024 C0702 "EOP Modifications"
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-_OD24 C0703 " Recovery from SFAS Actuation-
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0024'C0704 "AFW FC,LTBV,'&'ADV Local' Operation
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0024'C0713 "ICS Modifications"
0024 C0714 "IAS Modifications" .
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0D23 K0500(series) " Introduction to EFI"
To evaluate the licensee responses to significant operating events
the inspector reviewed the following documents related;to'the'
burnout of pressurizer heaters and the inadvertent. release of _ fuel
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pool leakage.
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Letter Ward to Martin, JEW 86-963, dated December 3,'1986
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Root Cause' Report 86-34 Pressurizer Heater Damage,-Draft dated
:1/30/87- '
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Op'erations Special. Orders
86-35'USE OF INFORMATION STICKERS 11/24/86
86-37 START.OF SHIFT CREW BRIEFING-11/26/86
86-34 RESPONSE TO' INDICATIONS >11/24/86
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Letter. Ward to Martin,' JEW 86-980, dated. December 15, 1986,_
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Nuclear Tracking Department Item, entry form 8700127 dated
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;1/14/87 ,
L Procedure-A-21 Spent Fuel Pool Cooling System,-Rev. 13 dated
$. 9/25/85 and Temporary' Change 8700002; dated 12/31/87- . The inspector observed;two days of B&W simulator training at
Lynchberg, Virginia to determine'the effectiveness of " hands-on"
training and to verify the actual training received.
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The following ~
evolutions were observed. .
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[ 0D24 K7900 "S/G Tube Rupture- l~_
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0024 K5500 " Loss of Off-site Power
"EFIC Operations"
- " Reactor / Plant Startup
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P ' IThejicensedandLnon-licensedoperations'personnelreceivedthe~~ ~
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same classroom'and required reading _ assignments. ~ The licensed'
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personnel;also received simulator training, annual oral and written
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examinations and required Technical. Specification 1 reading. 'The
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, ~.c lassroom training appeared to be professional and technically-
correct.-Questions which could'not be answered in. class were'
. subsequently investigated =and answered by the instructor
h The approved requalification program permits *up.to 30% of the. ' . lecture' series to be missed for a justifiable cause with make-up
% training, conducted within six weeks.of; return to duty. The
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, ; inspector.noted that on December 29, 1986'(Memo-NOS 86-345)'the
Shift Operations Superintendent documented instructions'to the '
t Shift Supervisors to. improve lecture' attendance. Shift Supervis'rs o ._
were directed that: (1) vacation requests.during training week 1
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be approved unless some sort of emergency exists, (2);the Shift
Supervisor must write a memo to the Shift,0perations-Supervisor
whenever an~ operator: misses a day'of training, stating the reason,
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weekly examination.
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Based on the sample examined,-the training ~ material and
presentation ~s related to.the December 26,- 1985 overcooling event- ,
i, appeared satisfactory. The training staff has repeated training on
manual valve operations to the' operators.since NRC inspection
- 50-312/86-07 was conducted. The training staff has also taught : t
, ' response to overcooling and_ Emergency Operations Procedures (EOPs), * 1
.and selected-systems at least twice since that time. The inspector
determined that this.was due (at least in part) to the lack of final
' design for the'EFIC. 'The design now appears fixed and final l
i. training has been scheduled for March 1987. The inspector alsoi < !
observed that the licensee had revised the E0Ps content and format.
This was done to verify adequacy, incorporate changes to the
facility, and to make the E0Ps more readable. ~
a , .At the time of the inspection, the training material did not cover
- '
the missunderstanding of pressurizer level instrument status which
resulted in burnout of'a=p'ressurizer heater (letter, Ward to Martin,
- . JEW 86-963, dated December 3, 1986) or the inadvertent release of
J. water from the fuel. pool leak chase system-(LER 86-25).-The
- ~ inspector questioned the training staff regarding training followup
of these two events since both events had occurred in' November 1986.
The training staff explained that they did not'. incorporate
- - operational events into the training material until an analysis or
Root Cause Analysis was finalized. They also explained that the
H[ Operations Department will normally issue Temporary changes to.
-
procedures and Operations Special Orders to keep all personnel
informed of these undesirable events. The inspector verified that'
these two events had been entered into the Training Department's
t followup system and assigned to a member,of the staff for final L followup. The' licensee training staff has scheduled the final ,
. Root-Cause. Analysis (86-34) of the pressurizer heater event and the'
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. spent. fuel pool leak (LER 86-25) to be included in the next required
reading; assignment. The inspector also verified that the licensee
management had issued Special Orders'and procedure changes-
consistent with the reported events and that the operators had read
these documents. Based on discussions with the training staff,
. operators, and.a review'of old training material the inspector
determined that previous training did not address these events.
The simulator training was conducted in two phases: first,
demonstration by the training staff and simulator instructors, and
second, unassisted operator plant start-up and response to
unannounced casulties. In the first phase, an instructcr
' demonstrated operations of the new and modified Rancho Seco systems
as well as control of plant pressure, temperature and level
following major plant upsets. The training stressed diagnosis of
plant conditions based.on all available instrumentation,
understanding of overall plant response to single and multiple
failures, implementation of a formal shift organization, and formal
communications between members of the. shift. During the second
phase, the operators were monitored by a minimum of three members of
,
the training staff during a plant start-up and during unannounced
l casulties. .After each session was completed the operators and
evaluators met separately to evaluate the problems and performance
which occurred. The training staff then asked the shift personnel
to present their conclusions, after which the training staff
presented their own findings and criticisms. The. inspector observed
that~the operators maintained a professional approach using formal
communications, adhered to written procedures, and properly
diagnosed plant conditions. The debrief sessions were succinct.
l
The scenarios p'repared by the licensee's staff clearly stated what
.
was' expected of each operator and correctly described the simulator
response.
The inspector concluded that the licensee management appeared
committed to take action to correct decreasing attendance at
lectures during training week before=it became an issue. The
written course material and classroom training appear to cover all
aspects of the overcooling event of December 26, 1985 and adequately
cover the local operation of plant valves. The current course
material appears to be well prepared and presented in a professional
manner. For.the three operational events examined it appeared that'
training before the events occurred was not sufficient to have
prevented or mitigated the events. 'However, previous classroom and '
simulator training appear to have met the requirements of the
approved requalification training program. The additional training
(on excessive cooldown,-loss of ICS, and manual valve operations)
after the overcooling event appears adequate to prevent similiar
future events. Training related to the pressurizer heater burnout
and fuel pool leakage had not been implemented as of the date of
inspection. '
The incorporation of operational events into the ongoing
Requalification Training Program appeared to lag the actual events
by approximatly four months. The initial information dissemination
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~ and corrective. actions; associated with~these events;are-promulgated
by the Operations Department.) 3This approach appeared satisfactory
; ,
. isince the' operators are informed of these events.promptly'and - ~
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subsequent retraining is required;after.the' event: analysis ~is _s .
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-Thksimulator.thaniniappearedtobiespeciallyeff4ctive.: The'
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' necessity for, and acceptance.of,' formal shift organization and'_
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communications by -training.and. operations personnel demonstrated '
clear training goals and detailed criteria for successful
-
. performance during simulator scenarios relat'ed,to'the 12/26/85
g event.
'
, 5. Interviews and Observations of On-Shift Pe'rsonnel-
, . . . -
To complete the assessment of training effectiveness and to verify
that the records examined reflected the actual training received,.
the inspector. conducted. interviews and plant walkthroughs with shift
operations personnel. This'was also done to determine if the
additional training given-to correct the deficiencies identified in
NRC report 50-312/86-07 (items 86-07-02 and 86-07-06)'had been
-
effective and to evaluate,the operators' knowledge related to_the-
pressurizer. heater. incident and unmonitored fuel pool-leakage-
-
~
incident. -The inspector's sample included twelve personnel from
three shifts and was divided between licensed and non-licensed-
operations personnel. The sample also included several of the.
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licensed personnel whose requalification records the inspector had- <
4 = examined.
~
'The inspector ~ questioned the operations personnel and requested-that_-
-
-they take the inspector to selected pieces of equipment. In'several
n cases the inspector required the operator to demonstrate. valve:
operation, trace. system flow paths, demonstrate appropriate
, ,
,
' procedures, and discuss the causes of the operational events
selected. The following general items were included.
^
,
Local ADV, TBV, FW, and AFW system valve' operations.
L- Local RCP Seal Injection and Seal Return valve operation L -Local Letdown Isolation valve operation l Aux. Steam Reducing station local' operation- '
Operation.of valve controllers in the control room
~
Operation of the make-up tank outlet valve during SFAS
Limitorque M0V. local operation
i- ! Local operation of Feedwater Heater valves
! , Position indication and failure modes for the valves described
- ,
above (loss of air, loss of power, and loss of control
l. signal)
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Identification and effects of complete and partial losses
of NNI and ICS power
Location of circuit breakers and indications for NNI and ICS
power supplies. ;
Location and limitations on' operation of Spent Fuel Pool Leal
Chase Channel isolation valves
Identification of inaccurate or out of service pressurizer
. level indication
Actions required following a pressurizer heater breaker trip
during heat-up
Radiological measures taken during an emergency entry into a
potentially high radiation area (potential fuel damage)
Recognition of an overcooling event
Action to be taken-during an overcooling event
Entry into known contaminated areas
Observation of control room supervision and communications
The inspector determined that, in general, all personnel interviewed
could find the requested plant components, understood local manual
valve operation, and had a good understanding of how the operable
installed valves worked.:The inspector observed that in the
situations where valve modifications were not complete some
operations personnel were. uncertain of what would be installed and
how'it would function.
However, two exceptions were noted. The inspector found that-the.
auxiliary steam reducing station was not fully installed and that
~
training appeared incomplete for this item. Some personnel were not
aware of how the new reducing station would operate, its failure
modes, or how.it could be controlled locally.
The inspector also found some confusion over the control system to
-
be used to control the atmospheric dump valves, how the back-up air
~
supply systems would be used for these components, and valve
position during~ normal operations. This appears to be due to the
fact that installation of these components is not complete and that
some aspects of the modifications had not been finalized as of the
date of the inspection.
The inspector found that some operations personnel were unaware of
any new controls placed on the isolation of the spent fuel pool leak
chase system. Some personnel were not-aware that an inadvertent
release had occured. The licensed operators questioned were aware
of administrative controls placed on these valves'and the
inadvertent release.
.
15
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Licensee training personnel responded that training in these two
areas was not yet completed.
The inspector observed that the s'hift personnel in the control room
were extremely busy during all three shifts. This activity in the
control room appeared to be due to ongoing work in the plant which
required constant attention to clearances and tests in progress.
The inspector noted that it was very difficult for the shift crew to
spare even one person for walk-throughs. The clearance coordinator
was often surrounded by several people and the senior licensed
operators appeared to be involved in many activities at the same
time. The inspector related this observation to the licensee's
management. The inspector stated that it appeared inappropriate for
scheduling and construction to control the plant operations to the
degree observed. The inspector stated that licensed operators must
be fully in charge of the plant at all times and should not be
subject to excessive pressures by other groups.
The inspector also stated that the knowledge and ability of the
operations personnel appeared satisfactory. However, since
extensive modifications had been made and some personnel stated that
they were unsure of what was going to be installed in the plant, it
appeared essential that the plant testing and start-up be (1)
deliberate and' controlled with no surprises, and (2) that all shifts
be involved in testing and start-up to familiarize everyone with the
modifications and new plant characteristics. A licensee
representative' stated.that they agreed that operations must be in
charge of the facility and would look into both issues.
The following week, the inspector observed that the operators
appeared to'be controlling the ongoing work at a more reasonable
pace and without the hectic pressure previously observed.
The inspector found that the operations. personnel examined appeared
!
to have received training consistent with_the documentation in their
records and the the training had been generally effective
l The inspector concluded that the additional training of licensed and l non-licensed operators was acceptable and that the deficiencies f identified during inspection 50-312/86-07 related to training should
be closed. NRC open items 86-07-02 and 86-07-06 are closed.
Information related to the uncontrolled release of fuel pool leakage
had not reached all of the non-licensed operations personnel.
Licensee personnel have stated that this material will be in the
second required reading assignment for 1987.
! Operations personnel appear to have been under pressure to allow
construction and testing to proceed as fast as possible. Training
personnel appear to have been under pressure to teach final system
l
design and operations just prior to start-up (system design freeze
l appears to have occured very recently, e.g. EFIC, Auxiliary steam). l The immediate effects of the inspector's preliminary comments to
licensee management appear to have been effective in asserting
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greater. operator control of the plant. However the inspector's
observations were limited and the effects on training and the plant.
e testing and start-up have not been observed.
No violations or' deviations were identified.
7. Followup of Unresolved Items
In inspection report 50-312/86-42 the inspector identified three
unresolved items which dealt with the licensee's apparent lack of
immediate notification to the NRC per the regt.irements in 10 CFR 50.72.
The licensee reviewed their control roon logs and provided the inspector
with a record of a red phone notification on September 7, 1985. This
phone notification discussed the events reported in LER 85-18. Therefore,
,
unresolved item 86-42-02 is closed.
However, the-licensee was unable to identify immediate notifications to
the NRC on the events surrounding LER 85-13 and 85-20.
The event discussed in LER 85-13, initiation of the emergency diesel
generator, appeared to have met the criteria in 10 CFR 50.72(b)(2)(ii) as
a four hour report which is "Any event or condition that results in
manual or automatic actuation of any Engineered Safety Feature (ESF)...".
The event discussed in LER 85-20, the HVAC system failure to function in
accordance with the design basis report and technical specification 4.10
requirement, appeared to have met the criteria in 10 CFR 50.72(b)(2)(i)
as a four hour report which is "Any event, found while the reactor is
shutdown, that, had it been found while the reactor was in operation,
would have resulted in the nuclear power plant, including its principal
safety barriers, being seriously degraded or being in an unanalyzed
condition that significantly compromises plant safety."
The apparent failure to notify the NRC per the requirements of 10 CFR
50.72 in two cases is an apparent violation, (Violation 87-06-03).
Unresolved items 86-42-01 and 86-42-03 are Closed due to the above
violation being issued. (87-06-03) Open, (86-42-01) Closed,(86-42-02)
Closed, and (86-42-03) Closed.
UNR 83-23-01, (Closed) " Feasibility of Analog Status of SFAS Added to the
Control Room.
In Inspection Report 83-23, the inspector identified an unresolved item
that in the Rancho Seco control room there were no annunciators of the
status of the three analog SFAS channels. The licensee committed to
review the feasibility of adding the analog status to the Integrated Data
and Display System (IDADS) computer.
In a memo dated September 3, 1986 from D. Gillispie to R. Ashley, the
licensee stated that their study determined that to add the analog status
of SFAS to IDADS would require additional equipment within the SFAS
system. Therefore, the licensee's Nuclear Engineering department, after
completing the feasibility study, recommended that the analog status not
m
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be added to IDADS. -This item does not, on further review, appear to have
been a potential violation,.and is closed. (83-23-01), Closed.
UNR 86-30-03 ,-(Closed) "Use'of Furmanite Leak Sealant.
The inspector reviewed'a. draft memo to be issued by Nuclear Engineering
(NE) which revised the NE guidance on the use of leak sealants
~
(Furmanite) for temporary repairs on safety related systems. The memo
revised a previous memo dated 1/2/87 and-limited the application of
Furmanite in safety related systems to temporarily repair existing
gasketing mechanisms in piping systems. Furthermore, it prohibited the
.
use of Furmanite for repair of through wall breaches in pressure
boundaries.
A licensee representative committed to provide by separate memo, a
current status of the applications of Furmanite in the plant and the
schedule for replacing any applications in safety related systems such
that prior to the next plant startup all known uses of Furmanite in
safety systems will be replaced with permanent piping repair.
In discussing this item with the inspector, licensee engineering
representatives identified that the clamp and sealant used for the repair
of thru-wall pipe leakage in the. drain line for the decay heat removal
pump were designed to appropriate ASME requirements as a mechanical joint
with gasketing material for leakage control. Furthermore, they contended
that the structural integrity of the original piping system was
unimpaired by the existence of the thru-wall defect and the addition of
the clamp. Consequently, the licensee concluded that the clamp and
sealant were properly designed to applicable code requirements and that
the original piping system remained in compliance with code requirements
despite the existence of the thru-wall defect and the addition of the
clamp.
In their engineering analysis, the licensee contended that the Furmanite
was not used as a " pressure boundary material" in the rr. pair, in the
sense that it was not analyzed as a replacement for the pipe wall
material. The licensee explained that their analysis was based on the
assumption that the thru-wall defect could be modeled as a pipe wall
penetration of known geometry. This assumption enabled the use of
established analysis techniques employing standard code calculations to
determine the acceptability of the pressure boundary. As such, the
licensee considered the design to be changed to incorporate a bolted
mechanical joint as part of the pressure boundary.
The lack of engineering control over the radiographic testing used to
identify and define the thru-wall defect geometry in support of the
engineering analysis has been identified in two violations in a previous
report. (86-18-02 and 86-18-03).
Based on their response to the previous violations and their programmatic
commitments limiting the future use of the leak sealant technique, it
appeared to the inspector that the licensee has taken appropriate
corrective actions in addressing the identified weaknesses in the
engineering program.
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1 In addition', .b'ased os followup.oflthis" item,[the following LERs were also
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LER 86-02-L1:' : Closed -
LER 86-14-LO: Closed c.
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8. ,-Licensee' Event Reports (LERs) Follow'up: +
7
. LER85-14 Revision 3 \)(Close)'-NonClass1ReplacementBreaker
'
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4- Installed,for NSCW Pump A. * ' '
'
The LER reported that a Class 2 breaker was use' d to replace a Class 12
0 breaker for NSCW pump A.while the' plant _was shutdown. The condition was' < '
.
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identified.on NCR 4796.on June 14,,1985. Reactor. criticality was
achieved.on June 14, 1985 and again on; June 23, 1985 without the NSCW "A" -
p trair being declared. inoperable andlwithout associated limiting i
- conditions-for operations being addressed. The condition was. recognized
- on July 9,1985 while the plant was again shutdown, appropriate Technical
Specification requirements were observed and a Class 1 breaker was
s ,
installed on July 24, 1985.
I
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Subsequent. licensee evaluation resulted in various corrective action -
, ' items, including establishment of a formal. review of NCRs-for impact on F
l . plant conditions and a formal program to document and control the ; j selection process for replacement parts.
The QA Department has performed several surveillances: documented on
"l surveillance report numbers ~746, 747, 758, and 766 to review completed <
corrective action for the' LER.. The corrective action for formalizing the
replacement parts control was in process but ha:i not yet been completed.
1 The NRC inspector determined that Operations Manual Procedure B.1, . l Revision 17, Plant Precritical Checks, hr/J been changed to include a- , ! requirement in Enclosure 5.5~ step 10 to verify resolution.of all NCRs as ! part of the precriticality check list. The NRC inspector also determine :
'
the Plant Manager and the QA Manager had issued internal policies for .
E issuance of safety related and environmentally qualified material from : ! the warehouse pending issuance of the revised replacement parts control i procedures. The policy requires.QC verification, tagging and control of I this material. The interim measures provided reasonable assurance that i the licensee's committed corrective actions will be implemented.
LER 85-14 was closed. The licensee's material iss'uance program will be
'
the-subject of future routine inspections.
i 1: LER 86-06, (Closed) 7 ,
The licensee reported that during cold shutdown conditions on
April 14, 1986, an ongoing comprehensive review of fire portection
- systems and procedures identified the licensee's failure to deluge fire
} suppression systems surveillances every three-years as required by the ! Technical Specifications.
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This' item is closed based on the adequacy of the licensee's corrective
actions,-which included the'following:
a. Performing the regnired surveillance testing
b. Cross-indexing of technical specification and surveillance
procedures to corresponding' surveillance test frequencies.
c. Revision of surveillance test procedures for deluge sprinkler fire
suppression systems to incorporate appropriate _ test criteria and
acceptance parameters as specified by governing code requirements.
d. Clearly outlining and dispositioning identified deficiencies through
nonconformance reports (NCRs). NCR #S-5627 was issued to track
deficiencies identified as a result of performing three year
surveillance _ testing of deluge sprinkler fire suppression systems
for fire zones 81 and 82.
LER 86-08-(Closed)-
The licensee reported that a proper fire watch had not been posted while ,
{
welding and cutting work was being perfo'rmed in fire areas. An hourly
fire watch was posted rather than a continuous fire watch
This item is closed based on the adequacy of the licensee's corrective
actions which included the following:
a. A commitment by the licensee stated in LER 86-08 to initiate
training for all fire protection group members the week of
May 25, 1986_which required completed review of applicable technical
specifications and administrative procedures for control of
combustibles and ignition sources in order to enhance the staff's
ability to abide by fire' protection technical specifications.
b. Maintenance of records of. staff: members trained and the training
lesson plans. e
c. Installation of fire detection for_ reduridant trains required for
safe shutdown in Zone 46 as outlined in' proposed Amendment No. 137
to the Technical Specifications.
~
The inspector also determined that several events between September, 1985
and January, 1986 had not been reported or were reported late,--
apparently:
a. Engineering discovery on January 7, 1986, that insufficient voltage
would be available to the essential control room HVAC power supplies
following a design basis LOCA, as required within 30 days after the
discovery of the event under.10 CFR 50.73(a)(i) and later identified
in LER 86-23.
b. On September 9, 1985, the licensee released liquid effluent without
on-line rad monitor or a dual verification of a chemical analysis of
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.the. release-sample.as req'uired by Tdchnical Specification 3'.15,' ,
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therefore,ithe event is reportable.under 10 CFR 50.73(a)(2)(1)(B).
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.c. While in a cold' shutdown condition ~ o'n September 10, 1985, the
s . licensee ~ declared both emergency diesel generators' inoperable. :This. l
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Jis a condition.that alone could have prevented the fulfillment of
the. safety function'to remove residual: heat and~was reportable under
, ,
10 CFR 50.73(a)(2)(v)(B). ,
, d. On October 30, 1985'; whil'eithepl'antLwas'inhotshutdown,the'
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auxiliary feedwater pump, P-319,' auto started on the safety.. feature
signal of low-feedwater pressure'. Therefore, this' event was an-
automatic actuation of an engineered' safety feature' system and
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reportableiunder 10 CFR 50.73(a)(2)(iv)., . .
"
, .
e .' 'On December.12,E1985,' the' Technical' Specifica' tion leak test', for? the
N, "
personnel hatch,'was performed.without maintaining technical
specification. required pressure...! This condition is prohibited by=
: Technical Specifications, and therefore_ reportable under 10 CFR
50.73(a)(2)(1)(B). ,
c* f. On September 23, 1985 the decay heat "dropline"/ suction isolation'
,
valve closed spuriously, removing the ability of the system to .
, remove residual heat from the valve. - This condition was reportable
, under 10 CFR 50.73(a)(2)(v) but was not reported. x
t-
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-This'is an apparent violation of 10 CFR 50.73 (Violation 87-06-07). .
8a.' ?Special Reports Followup
LSpecial Report 81-59-X0, (Closed)
a,
The111censee reported an hourly ~ fire watch had been posted instead of a
continuous fire watch as required by Technical Specification
m. No.- 3.14.6.'2.> .
ThisJitem is closed based on the adequacy of the licensee's corrective
actions which included the following:
a. Revision of Administrative Procedure (AP.) No. 60,
Section 4.3.1.1.3.5 to clearly identify specific zones which require
continuous surveillance based on Technical Specification
a ' requirements for areas containing equipment important to' safety.
'
b. Implementation of an Administrative Procedure review to ensure that
sufficient emphasis is placed on the referenced Technical
: Specification requirements.
.Special Report 82-01-X0, (0 pen)
.
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.The licensee reported a fire watch which was not performed for three
- hours due to airborne radioactivity.
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This~ item remains'open pending' final resolution by the licensee.
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Resolution.to this problem is in progress. However, adequate
administrative controls were.not in place during the inspection to as'sure.
that minimum qualified staffing'will be:on site at all times to perform
,
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fthis duty.
Special Report 82-02-X0, (Closed)
~
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The licensee reported that an hourly fire watch in the "B" Diesel
Generator Room'was.not conducted for four hours.
This item is closed based on the adequacy of the licensee's corrective
actions taken.-This was an unusual occurrence. A fire watch was required
in the "B" Diesel Generator Room because of a non-functional' fire barrier
penetration seal. The door to the Diesel Generator Room had been.
replaced with a new door at 3:30 p.m. ' Inadvertently, a cover plate' was
rot . installed over the normal door latching mechanism so that the door
would be controlled only by it's security latching device. . Because the- .
plate wasLnot installed, as required, when the door was closed it latched
in a locked position. When the fire watch attempted to gain entry into.
the room, the security latch _ opened but the door latch held the door
closed. No key was available for the door latch immediately,.so it took-
.until 7:24 p.m. to gain entry into the, room without causing damage to the
door. Normal fire detection and suppression equipment in the room was
verified operable during the period of time.that the fire watch was
unable to gain entry into the room. =The required cover plate was
installed over the normal ' door -latching mechanism and= the key for the
door was made available to security and the control room.
~
Special Report 83-01-X0,:(0 pen)
The licensee reported the reactor' coolant pump room fire detectors were
not tested as required by Technical-Specification No. 3.14.1.
The licensee's surveillance procedure no. SP-201.03W was being revised to
address this problem (accessible vs. inaccessible testing). -This item
remains open pending further licensee action and NRC review.
Special Report 83-02-X0, (Closed)
The licensee reported the motor driven fire pump out of service over
seven days for maintenance.
This item is closed based on the adequacy of the licensee's corrective
actions. During the March, 1983 outage, Motor Driven Fire Pump No. P-440
was administratively taken out of service for more than seven days due to
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the necessity to drain the cooling tower circulating water basin. A
recovery procedure was established for the pump which required filling
the basin by opening inlet bypass valve No. PCW-002 from the Plant
Cooling Water System to provide sufficient water supply for the pump to
be manually operated as needed. The recovery procedure remained in place
until the Cooling Tower Circulating Water Basin was refilled. The Diesel
,
Driven Fire Pump was operable to provide automatic flow to the Fire
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. , ~ Protection.Wate(Supply;SystemduringthE: entire-periodof.timethatthe
s. Motor Driven Fire Pump wasideclared; administratively inoperable.
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. ,Special' Report 83-03-X0,=(Closed)
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s The 11censee reported the unavailability of fire su'ppression wat5r-
supplies. .
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This item is. closed based lon the licensee's correctiv'e^ actions 1taken a's
, s ' discussed in'the special. report. 4 . ,
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y Special Report 83 204-X0,:(Closed)
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Licenseereportedunavailability'ofitwo,highpressure~ fire l pumps.'
This item is closed based onthe licensee's correstive actions'taken as
discussed in the~special report. ,
'
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Special Report 87-01-X0, (0 pen)~ .
~~
a- The licensed reported non-functionalffire barriers required to be ,
'- operable by Technical Specification 3.14.6. y,
;This report 'is a periodic update of previous- special reports addressing
the same s'ubject.- Report No. 87-01-X0 indicates that'the licensee is
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making progress toward making breached fire barriers operable. .According.
, , -to the licensee, prior to startup from the current outage, the licensee's
goal is to have all breached fire barriers operable. The large number of'
breached-barriers may-be'significant',-and this item remains open pending
further licensee action and NRC review.-
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'9. Inspector Followup Items-
'
:Open Item (50-312/85-22-01), (C1osed).
By. letter dated February 28, 1985, the licensee submitted a formal-
. exemption request to the NRC for relief.from the requirements for
. protection offredundant safe shutdown trains in accordance with
: , Section III.G.2 of, Appendix R.to 10 CFR 50 for' Fire Area No. 69,
- Auxiliary Feedwater Pumps per the provisions'of 10 CFR 50.12.
This item is closed based on the NRC's review of the. requested exemption
which will be discussed in.a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
.
Open Item (50-312/85-22-02),-(Closed)
By letter dated February 28, 1985, the licensee submitted a formal
exemption request to the NRC:for relief from the requirements for
protection of redundant safe shutdown trains in accordance with
Section III.G.2 of Appendix R to 10 CFR 50 for Fire Area No. 69,
auxiliary feedwater pump valves, per the provisions of 10 CFR 50.12.
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This item is closed < based on the NRC's review of the requested exemption
'
which will be discussed in a supplemental Safety Evaluation Report to be
, issued;by.the NRC at a later date.
Open Item (50-312/85-22-03) '(Closed)' ' '
~
By letter dated February 28,~1985,;the licensee submitted a formal-
exemption' request to the NRC for relief from the requirements for
protection of redundant safe shutdown trains in accordance with
'
Section III.G.2 of. Appendix R~to.10 CFR 50 for Fire Area No. 68 cables
inside' containment electrical' penetration area per the provisions of
~
10 CFR 50.12.
This item is closed based'on the NRC's review of the requested exemption
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-22-04), (Closed)
By letter dated May 4, 1985, the licensee submitted a ' formal exemption
request to the NRC for relief from the requirements for protection of
redundant safe shutdown trains in accordance with Section III.G.2 of
Appendix R to 10 CFR 50 for Fire Area No. 69 Nuclear Service Cooling
Water Pumps per the provisions of 10 CFR 50.12.
This item is closed based on the NRC's review of the requested exemption
which will be discussed in a Supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-22-05), (Closed)
By letter dated February 28, 1985, the licensee submitted a formal
.
exemption request to the NRC'for relief from the requirements for
protection of redundant safe shutdown trains in accordance with
Section III.G.2 of Appendix R to 10 CFR 50 for Fire Area No..RG1 cables
inside Auxiliary Building Electrical Penetration Area per the provisions
of 10 CFR 50.12.
This item is closed based on'the NRC's review of the requested exemption
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-22-06), (Closed)'
By letter dated February 28,.1985, the licensee submitted a formal
exemption request to the NRC for relief from the requirements for
. protection of redundant safe shutdown trains in.accordance with
Section III.G.2 of Appendix R to 10 CFR 50 for Fire Area RB1 Auxiliary
Building Mechanical Penetration Area per the provisions of 10 CFR 50.12.
This item is closed based on the NRC's' review of the requested exemption
.
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
.
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Open Item (50-312/85-22-08), (Closed)
By letter dated. February 28, 1985, the licensee submitted a formal
request to the NRC for relief from the requirements for protection of
redundant safe shutdown trains in accordance with Section III.G.2 of
Appendix R to 10 CFR 50 for Fire Area No. 110 Nuclear Service Raw Water
Pumps, per 10 CFR 50.12.
This item is closed based on the NRC's review of the requested exemption
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date
Open Itiem (50-312/85-22-09), (Closed)
By letter dated April 4,1985, the licensee submitted a formal exemption
request to the NRC for relief from the requirements for protection of
redundant safe shutdown trains in accordance with section III.B.2 of
Appendix R to 10 CFR 50 for Fire Area 91, Nuclear Service Electrical
Building roof, per the provisions of 10 CFR 50.12.
This item is closed based on the NRC's review ~of the requested exemption
which will be' discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a-later date. '
Open Item (50-312/85-22-10), (Closed)
~
'
By letter dated May 24,1985,-the licensee submitted a formal exemption
request from the requirements for protection of redundant safe shutdown
trains in accordance with Section III.G.2 of Appendix R to 10 CFR 50 for
Fire Area No. 74 Auxiliary Building roof Control Room HVAC units per the
provision of 10 CFR 50.12.
This item is closed based on the NRC's review of the requested exemption
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-22-11), (Closed)
By letter dated May 24, 1985, the licensee submitted a formal exemption
request to the NRC for relief from the requirements for protection of
redundant safe shutdown trains in accordance with Section III.G.2 of
Appendix R to 10 CFR 50 for Fire Area No. 74 Auxiliary Building Roof
Nuclear Service Cooling Water Surge Tank level switches per the
provisions of 10 CFR 50.12.
This item is closed based on the NRC's review of the requested exemption
which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-22-12), (Closed)
By letter dated November 7, 1985, the licensee submitted a formal
exemption request to the NRC for relief from the requirements for
protection of redundant safe shutdown trains in accordance with
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SectionlIII.G.2ofAppendixRto10CFR50fortheControlRoomFire
Area 1 per the provisions of 10 CFR 50.12.
~
This item.is closed based on the NRC's review of the requested exemption
-which will be discussed in a supplemental Safety Evaluation Report to be
.
issued by the NRC at a later date.
Open Item (50-312/85-22-13), (Closed)
By letter dated February 28,1985, the licensee submitted a formal
exemption request to the NRC for relief from the requirements for
capacity of Reactor Coolant Pump oil collection systems in accordance
1with Section III.0 of Appendix R to 10 CFR 50 per the provisions of
,
,10-CFR 50.12.
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This: item is closed based on the.NRC's review of the requested exemption
_ which will be discussed in a supplemental Safety Evaluation Report to be
issued by the NRC at a later date.
Open Item (50-312/85-36-01), (0 pen)
The licensee's administrative procedure review and programmatic
assessment of effective implementation of fire protection program
-requirements is ongoing. This item' remains open'pending completion of
licensee actions and further NRC review.
'
,
Open Item (50-312/86-18-07), (0 pen).
This item concerned apparent lack ~ ofd ' rainage capability for discharged
- automatic fire suppression water in the technical support center
~
The licensee'is performing an' engineering study to determine potential
adverse affects from this condition. According to the licensee, this
assessment is ongoing. This item remains open pending completion of
~ licensee action and.further NRC review.
.
Followup (50-312/85-11-01 and 85-27-01), (Clos'ed)
Inspector Followup Item 85-11-01:noted five inspector concerns regarding
the licensee's QA program procedures. Inspection Report 50-312/85-27
reviewedLand closed out the five concerns of followup item 85-11-01 but
opened followup items.85-27-01 and 85-27-02 regarding additional concerns
with QA program procedures. Inspection Report 50-312/85-31 reopened
followup item 85-11-01 pending issuance of a revised QA manual to
specifically address nuclear operations.
The NRC inspector reviewed the QA manual and determined that on
, January 1, 1986, the QA manual procedures had been changed and QA policy
sections had been issued to include QA program requirements for the
operations phase. Inspector Followup Item 85-11-01 was Closed.
Inspector Followup Item 85-27-01 had been opened pending issuance of a
formal review process for 10 CFR 50.54(a)(3) QA program changes. The
inspector reviewed Amendment 17 to the Management Safety Review Committee
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(MSRC) Chapter and determined that proposed changes to the QA program
description which involves reduction in program commitments previously
accepted by the NRC, were added to the required MSRC review process.
Inspector Followup Item 85-27-01 was Closed. The NRC inspector also
determined that the associated Followup Item, 85-27-02, had been
previously reviewed by Region V and was Closed on Inspection Report
50-312/86-38.
Followup on Deviation 86-07-10 (0 pen) " Control Cable Shielding Not
Protected at Ungrounded End"
In response to this deviation, the licensee committed in a letter to the
NRC dated June 13, 1986, to either perform a sample inspection of the
terminations in the plant and to rework as needed or to use analysis to
demonstrate that with the extensive ground grid system, the identified
discrepancy is not a genuine technical concern and that no further
inspection and rework is necessary.
Subsequent to the above letter, the licensee. contracted Bechtel Power
Corporation to perform the analysis. In Bechtel's Letter No. BSL-6160 to
the licensee dated 10/7/86, Bechtel concluded that " Incorrect grounding
of instrument cable shields and signal conductors creates ground loops
that may cause degraded analog signals when electrical power system
ground faults or lightning occur....Because of the.brief duration of the
degraded signals, only circuits that have memory function to retain the
signal action can cause unacceptable consequences." The unacceptable
consequences include:
A. Initiate a unit trip
B. Initiate a safety feature actuation
C. Inhibit a safety features actuation
D. Change the state of major plant equipment
E. Upset major plant-process control devices.
F. Prevent plant operators from performing essential functions.
Bechtel identified a number of' plant systems where degraded signals
caused by incorrect instrument cable shield and signal conductor
grounding may inhibit operation of plant safety systems or cause
unacceptable upsets in plant operation. These systems are both Class 1E
and Non-Class 1E such as:
A. Reactor Protection
B. Reactor Control
C. Nuclear Instrumentation
D. Safety Features Actuation
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, EightyL(80). cables of4 these systems were identified by Bechtelifor
l reinspection walkdown-and rework,;if the ground shield terminations were
found not. insulated. -The licensee 1 initiated work requests and, included .
thenLin the work planning for implementation ~ prior to the restart.
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-Theinspectorconsideredthelicensee'scorrectiveactionstobe
responsive and will follow the implementation during future inspections.
This. item remained open pending licensee's completion of the walkdown and
'
necessary rework. " , ,_ !
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LER 86-10, LER'85-16,'Rev'.1 Cable Raceway Trackina Systems'(CRTS),-(0 pen)
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1. ' Introduction ~
(
During this inspection period, the-inspector continuously menitored
the licensee's corrective efforts in this area by in office review
and onsite inspection. ? Previous' inspection results were documented
in Inspection Report 50-312/87-01. This. inspection will be -
continued, and> remains open pending completion of the licensee's
program and. review by the NRC.
The scope of the licensee's resol _ution of the CRTS problems included
the disposition of the discrepancy generated by the-CRTS.
self-checking features 'and the discrepancy between the CRTS data
,
base and the actual plant configurations. ' .
,
2. CRTS Documentation Discrepancy Resolution-
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The licensee stated that when the original Bechtel's' cable tracking
~
. system EE 553 was converted to the SMUD CRTS in 1981, a
self-checking program was run on the~ CRTS. and it identified a number
'of discrepancies in meeting the CRTS built-in criteria. These
discrepancies' included cable trays filled beyond the design
criteria, mixing of instrument cables with power and control ' cables
in the same raceway, mixing of class 1 and non-class 1 cables, and
missing nodes where cable raceways should be physically connected.
The licensee was dispositioning these discrepancies by reviewing
against the Updated Safety Analysis Report (USAR) Chapter 8' '
Commitments, Appendix R and SMUD current construction standard. l
A. Mixing of Instrument Cables with Power / Control Cables
The licensee reviewed approximately 350 instances of instrument
and power / control cable intermixing and did not identify any-
cables needing rework due to misrouting. The discrepancies
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were. accepted by the licensee based on one of the fol'owing
criteria:
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(1) The cable carried a digital signal which was not
susceptible to external noise.
(2) The instrument power supply cable and signal cable shared
the same connector or raceway while the instrument power
supply cable carried a constant current.
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(3) The' instrument was used only for indication or alarm
function without control or actuation functions. External
noise would result only in momentary change of indication.
However, because the USAR did not specifically provide-
allowance for these criteria, the licensee is reevaluating
whether rework is necessary.
B. Cable Tray Overfill
The licensee's USAR commitment stated that the maximum fill in
redundant trays is 40 percent. The CRTS self-checking run
identified 38 class 1 and 150 non-class.1 cases of overfills.
The licensee dispositioned these overfills by evaluating the
current carrying capacity of the cables in the trays and
structural loading of these trays. The licensee identified one
case of overfilled tray exceeded the 50 pound per linear foot
limit by 0.2 pound . The licensee notified the NRC per
10 CFR 50.72. Currently, the licensee was performing seismic
structural ~ analysis of this finoing. Furthermore, the licensee
was evaluating other power cable tray's less than the 40 percent
filled to ascertain whether the criteria was conservative
enough.
C. Mixing of Class 1 and Non-Class 1 Cables
Cable raceways with this condition were identified by the CRTS
self-checking due to the generic application nature of the
vendor. supplied software. However, the Rancho Seco USAR
permits this condition provided that the non-class 1 cable does
not form a bridge between two redundant class 1 raceways. The
I licensee did not identify any condition which violated this
USAR commitment and required rework.
l
D. Cable Raceway Connections
One of the features of the CRTS self-checking program is to
check all the raceways.which a cable is supposed to go through
are connected in the order of the cable routing. The raceway
configuration is part of the CRTS data base. Approximately 200
cases of class 1 and class 2 required by appendix R were
,
identified by the CRTS self-checking run to have raceways not
i
connected according to the routing. Typically for these
discrepancies, the CRTS was unable to locate the node of
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connectionof[thetwo' trays-whichacable'was.supposedtogo
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through'. Theilicensee completed the resoluted of these
discrepancies by reviewing the cable routing schemes, raceway-
configuration drawings, and-in about 75 percent of the cases,_
: conducted. actual walkdowns of the raceways to identify raceways
~
-and connections not entered;into the CRTS routing schemes but- '
physically. existed between the two disconnected raceways. Upon
completion of these resolutions, the licensee updated the CRTS
routing schemes by, adding the missing connection to-the data
*
base.' The licensee did not' identify any necessary rework to
the cable or raceway as.a result of the self-checking program
g cases. -
The licensee's effort in resolving the discrepancies generated by
the CRTS self-checking program was nearly completed for class 1 and
'
for class 2 cables required by Appendix R was nearly completed also.
The licensee stated that none of the deficiencies were safety
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significant.' The inspector reviewed samples of the licensee's
disposition packages and ' independently walked down two package's and
found.them to be satisfactory. Based on these samples, the
licensee's' program to resolve these discrepancies appeared adequate.
~
However, 10 CFR 50.59 required the' licensee to perform a safety
analysis when a modification results in a condition not described in
the USAR (such as overfilled _~ trays and mixing instrument cable with
power / control cables). Since the licensee was unable to present
documentation of such a' review for these modifications, this is an
unresolved item (87-06-08), pending further review of records by the
licensee and the inspector.-
3. Sample Walkdown of~the Cables-to. Verify CRTS Accuracy
LER 86-10 revealed a discrepancy between the CRTS_ data base and the
actual plant configuration. To address the concern over the quality
and accuracy of the data base,-the licensee initiated a sample
'
walkdown program to' physically verify that the cables were actually
routed as the CRTS data base shown. The sample walkdown program
covers those cables installed or modified after the> plant was
constructed and turned over by Bechtel. The licensee believed that
the original Bechtel work was covered by an accepted Bechtel QA
program (this remains to'be verified by the licensee). This remains
to;be verified.
~
Initially, the_ licensee divided the cables into three lots:
o < Lot-1: Cables installed with subsequent modifications. The
licensee believed that this category was vulnerable to errors
, due to revisions not properly implemented;
o - Lot-2: ' Cables installed without ' subsequent- modifications.
o Lot-3: Cables installed under the responsibility of an
individual-construction engineer (CE) associated with the ~
modification reported in LER 86-10 which were not completed.
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The licensee initially' based th'e sample size on MILLStandard'105D to
achieve 95 percent confidence that upon completion of the sample the
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. rest of the lot'is accurate with a 95~ percent qualityllevel. _Later,
c the licensee switched _to a different model called " likelihood- j: -
1d ensity function" ~which the licensee' believed could maintain the-
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same' confidence and'_ quality level while reducing.the sample size.'
In the process of the sample walkdown, onl February 4,1987,.the^
licensee found_one. cable in Lot-3 sample' space-and one cable in
". . Lot-1 sample space were not routed ~ascthe CRTS data base shown. Th'e
,
second cable was also in the population space lof' Lot-3'. Further .
evaluation by the licensee revealed that these;two cables were
_ , installed under the responsibility of the'same individual along with---
five other~ cables which were also found misrouted. The as found
condition of these1.seven misrouted cables presented an Appendix R'
. violation since redundant signal cables needed for safe shutdown
~ .
. were in the same fire area. ;The licensee promptly took compensatory
, measures by posting a' fire watch and notified the NRC per
10CFR50.72. These seven cables-were then~ rerouted properly'as
_
. verified.by the inspector's independent walkdown. The licensee
'
v~ dete'rmined that these seven cables were in the same work package
ECN4942 as the other seven cables. reported in LER 86-10. tThese two
sets of seven cables. served the following: indication, functions:
-
o Steam Generator A Level <
. .
o Steam Generator B Level
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o Loop A RCS Cold Leg Temperature
s .o- Loop B RCS Cold Leg. Temperature-
o Pressurizer Level '
-o Make-UpTdnkLevel' l
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'o 1 Steam Generator Pressure ,
Thefirst~setlofcables'(LE[86-10)2were routed from an isolation
cabinet H4SIA to cabinett H4 CAL;in ~the control room and the second
set of cables (found 'during. the walkdown), were outed from H4SIA to
the Remote Shutdown Panel H2SD in the same rim
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Based on the above ' finding, the:licenthe nsp ted that the
misrouted problem most .likely occurrel y w 4 ECN was revised.
subsequent to the first' installation and.the quality of the work by
the individual responsible-for._the 14 misrouted cables. Thus , ~ the
licensee regrouped the lots.as follows:
;.
'
o ' Lot-1 'riginal~ Lot-1 minus those installed under -
the responsiblity of the associated CE. The licensee
selected a_ sample size of 56 out of the population
- size of.1,702.
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'the responsibility of the associated CE. _The
, licensee. selected a sample'siz'e of.52 out of the
. population size of 424. '
o LLot-3 Cable installed under the responsibility of
the associated CE without' subsequent revisions. . The.
licensee selected a sample; size of 48'out of the
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population size of-189.
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s o- : Lot-4 : Cable' installed under the responsibility of
:theLassociated CE with subsequent revisions. The
' licensee decided on a 100 ' percent inspection'of the
populati.on size of 78.~
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The adequacy 'of the sample' lot selection and sample size was
/. discussed with 'the office of Nuclear Reactor Regulation (NRR) staff. '
' The license'e committed to assemble a package with detailed
4 description of the sampling model for further, evaluation by the
staff. This issu~e will remain'open pending further NRC review.
_
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The physical walkdown of the selected cables were being. performed
under the-licensee's;Special Test' Procedure!STP-972 " Verification of
Instrument Cable Routing" and STP-976 " Verification of Power and
~
Control-Cable Routing."-:With the few exceptions of obvious short
, cable routings where visual inspections.were' adequate, the licensee
-utilized ~two radio. frequency traces to-aid.the cable tracing.
^
To assess'the-adequacy!of'the license'e's; effort, the inspector
conducted the following assessment <
o Reviewed the aboveitwo procedures.
,
o Reviewed samples;of the completed packages.
-
-o- LWitnessed .a sample' of two walkdowns in process with each of the
two RF tracers.
o Witnessed. a -visual . inspection ~ walkdown in process.
o Requested the licnesee to retrace two cables with the
inspector.
,
The cables sampled by:the inspector were routed through various
buildings of~the plant from the roof of the building to manholes s
underground, and from the' clean technical support center to.the
contaminated areas in the. plant.- Various types of raceways were
covered by the inspection. Based on these assessments, the
inspector found the licensee's performance of walkdown to be
-
reliable to confirm the exact locations of the cables sampled.
Near the end of the inspection, the licensee had completed
approximately 150 walkdowns. With the exception ~of the major
deficiency of the 7 cables described above, the licensee did not
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identify othe_r major defects. LThe licensee identified a few minor
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differences between.the CRTS and actual cable installation and
dispositioned them acceptable ~after evaluation against requirements-
such as the USAR and 10 CFR 50 Appendix R. The inspector reviewed
. .the licensee's dispositions and found them adequate. LER 86-10 will
-remain open pending completion of the CRTS program and the review by
,
the NRC.
4. (0 pen) LER 85-16, Rev. 1, " Spurious Closure of DHR Dropline
Isolation Valve" ,
~In addition to the issues discussed above, the inspector developed a
new concern from this LER, which had previously been closed in
Inspection Report 50-312/86-42. The licensee determined the root
'
cause of this LER was that the instrument cables for the pressure
transmitter PT-21099 was misrouted in the power cable trays and
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resulted spurious actuation of the isolation valve due to electrical
noise. The licensee stated that this cable was installed by Bechtel
prior to the turnover in 1975. The apparent cause of the misrouting
was that the design engineer specified the cable to be shielded
twisted pair and code the cable type accordingly. It was expected
that the cable routing engineer discern from the cable code that it
was an instrument cable and not to be routed with power cable. This
misrouting was not detected by the original Bechtel's cable tracking
system EE553 because it check the consistency of the cable trays for
their service level (power or instrument) but did not check between
the cable and the tray. When the EE553 was converted to CRTS, the
cable service type (power or instrument) was retroactively coded
based on the service level of the tray it was in. Thus this cable
passed the CRTS self-checking program undetected.
This deficiency raised the concern over the quality of the cable
installed by Bechtel which currently is not being addressed by the
licensee's sample walkdown program. The licensee committed to
prepare justification prior to restart.
10. Allegation Followup
RV-86-A-022 (Closed)
A. Characterization
Maintenance personnel are being blamed for operations errors.
B. Implied Significance to Operation
The wrong department or organization could potentially be singled
out for inappropriate performance, therefore, the corrective actions
taken would:not affect the people responsible for the occurrence.
C. Assessment of Safety Significance
The inspector reviewed the licensee's Licensee Event Reports (LERs)
l' for 1985. The inspector focused on LERs which involved safety !
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system actuations. :Five LERs were identified for review: 85-11,
85-15, 85-21, 85-23, and 85-25. . The licensee's discussion of the
root causes of these above LERs indicated that in three of the
events the. operator',s inattention or error; contributed to the event.-
The other two~ events consisted of: A contract electrician caused a
feeder breaker to open accidentally, and an electrical technician
error in lifting a' lead incorrectly. . 4
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D. Staff Position 4.
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The inspector:foundo 'nly two out'of the five cases of safety system
factuation reviewed involved the' licensee's acknowledgment that
maintenance personnel _ errors.were~ contributing ~ factors in the event.
In both cases the inspector concluded that'the licensee had
appropriately identified the personnel error. Therefore, in the
review the inspector did not identify any circumstance where-the
licensee had blamed the maintenance personnel for operations errors,
and that theilicensee had reasonably' identified the personnel
errors. .The allegation was not substantiated, and is closed.
E. Action Required: _
None
RV-86-A-078
A. Characterization
a. No procedure exists for acceptance of material and for
controlling nonconforming material found in the warehouse.
b. 0-rings have been issued for service that have had expired
shelf life.
B. -Implied Significance to Operation
a .' Safety related work is not being controlled by appropriate
. approved procedures.
b Unqualified, or nonconforming, 0-rings could be used in safety
related equipment.
C. Assessment of Safety Significance
The alleger brought both of these allegations to the attention of
the licensee's Quality Assurance Manager on the same day the
allegation was identified to the NRC inspector. The licensee's
conclusions from its investigation were that: "0-rings with expired
shelf life were withdrawn from the warehouse stock and were
processed and controlled in accordance with site procedures. Some
of these withdrawals were for the purpose of removing them from
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stock and triggering the system to order replacements. Others were
for use in non-safety, secondary plant applications. In no case
were any rubber components with expired shelf life installed in
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Quality /C1' ass'11 equipment without a Nonconforming Report-being
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;origi_nated and properly-dispositioned".1.(Excerpted from meno:
c i W. Koepke to S. Knight, dated.2/2/87) '
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1, ;Theinspectorreviewed[theilicendee's-investigatlionsonthe
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allegation, and d,evelope, d the~ following y
concerns:
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a. Thelicense[only" loded ~;at0-rings;thatwereissuedduringthe--.-
period of July- 14 through September 18,:1986. The review did-
not look at any 0-rings that.were -issued' prior to that period.
Th'e licensee's rationale for: limiting the< time-period-reviewed ' '
was that this'was~the interval during which the alleger was
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. involved on site with"0-ring storage,
b. The licensee stated that in_no' case were any rubber; components
with expired shelf life installed in Quality Class l' equipment
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~without a Nonconforming Report being originated and
dispositioned. The inspector reviewed the list of 0-rings
issued during'the period of July <14 through September 18, 1986
. .with a licensee representative, and the representative was
unable ~to definitely identify which 0 rings were issued to
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Quality Class-1 equipment. 'In addition, the l licensee was -
unable to identify the amount of time that the issued 0-rings
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had exceeded their shelf life.
c. .The Nonconforming Report (NCR) referenced above-is NCR 5806,
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issued 8/6/86. The disposition of the NCR required an '
engineering evaluation to be performed on expired items prior
to withdrawal of the item from the warehouse if the item's.
. shelf life had expired. The evaluation ~would encompass a
physical condition evaluation'and an evaluation of the storage ~-
environment. This evaluation would a' 6 be' documented on a
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. Receiving Inspection Data Report (RIDh,..
The inspector found that the disposition of the NCR was based
on a documented telephoneEconversation with a representative-
from a local supplier of 0-rings. The representative related -
that 0-rings could be stored for an extended interval provided
that temperature and humidity wer.e' monitored. No specific
interval or criteria were provided.' There was no other
technical information from the'various vendors which supplied
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rubber parts to the licensee to support the NCR disposition.
In addition, the licensee is presently not storing the rubber
parts'in the various warehouses on site as described in the
telephone conversation.
d. The inspector reviewed a sample of six.0-rings with stock
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numbers which the alleger had' identified as 0-rings with-
expired shelf life. Of thes~e 0-rings, five of six Sad
designations ~as non-Class 1 (non ' safety-related) parts.
Therefore, these items were'not< intended,~ at present, for
Class _1 equipment.' The remaining.0-ring has been in review for
Class 1 designation si.nce 7/30/86. The inspector found-
acceptance tags on 0-rings with,the above stock number,
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howeversit was_ not segregat'ed from 0-rings of similar stock /
numbers which had expired shelf life.
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f 4. Staff PositionT
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Thelal' legations we're' partially-. substantiated.' Th'e licensee's review ~of' . .
11 1t he allegations appeared to be incomplete as noted above. -Two violations - !
of. regulatory requirements have:been identified-(see paragraph 9). 'The.
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licensee will be required to respond,to these violations within.30 days.
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This allegation ~.is closed. Lic~ensee' corrective action to the yiolations; -;
n' will.be followed up in a subsequent inspection to ensure necessary
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-program changes are made. The itemsJin paragraphs'3a through d will be-
l- followed up.in conjunction with Violation 87-06-05 (Paragraph 9).
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Allegation RV-86-A-0099- .
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1. Characterization ;
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"QA memorandum SRK 86-079, dated November 17,'1986 tells QC inspectors ._
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not to write NCR's and their supervision will determine what NCR's are to
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.be written." ,
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, 2. . Implied Significance to Design, Construction-or Operation
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. QC inspectors are beingfdiscduraged,from writing nonconformance reports' ,
. by QA memorandum SRK.86-079. ./ .
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3. Assessment of-Safety Significance ,- -
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F Thefollowingprocedures.wepe' reviewed-toideterminethelicensee's -
nonconforming. condition reporting .-
requirements. ,
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QA Policy Section XV, Revision 0, Nondonforming Materials Parts or
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QA Procedure (QAP) Number.17, Revision 4', Nonconforming Material
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j '; Administrative sProcedure 22,' Revision ~ 12,' Occurrence Description '
Reports (0DR's) Reporting and Resolution
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nonconforming conditions. However the inspector had the following '
concerns regarding QAP 17: -
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(a) QAP 17 does not address potential 10 CFR Part 21~and 10 CFR Part
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50.73 reporting requirements. ,
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) [.' (b) General Requireme'nt paragraph 6.3 of QAP 17 allows the use of a
5-day memo for non-operational nonconforming conditions in lieu of a
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nonconforming > report (NCR). QAP-17 does not provide any controls or
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s . requirements (record keeping, close-out controls etc.) for 5-day
memo's.
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-(c) QAP 17 procedure paragraph 2.1 allows NCR's that are determined to
describe a condition-that was not a nonconforming condition', to be.
returned to the originator.without a' control number. No auditable
. record of returned NCR's is required.
-(d) 'QAP 17 procedure paragraph 2.6 requires a copy'of all "Operctional
NCRs" to be sent-to the Unit Operations Superintendent for review
and action to assure operations in compliance _with license
requirements,,however, QAP 17.does'.not require immediate NCR
notification of the Unit Operations Superintendent when the plant is-
E operating to assure prompt license compliance.
(e) 'QAP 17 procedure paragraph 3.1 requires NCR disposition _ action to
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include determination of cause and action to preclude recurrence but
does not consider potential generic nonconforming condition
evaluation for significant conditions adverse to quality.
(f) QAP 17 procedure paragraph 13.0 allows NCR's with control numbers to
be voided upon review and approval by the Engineering Review Board.
No criterion is provided for ERB approval to void the NCR.
The above noted QAP 17 concerns were discussed with the licensee, pending
further-review, the concerns were identified as Inspector Follow-up Item
87-06-09, "QAP-17 concerns". The licensee QA Manager stated that these
concerns would be considered for possible future incorporation into
QAP-17.
Memorandum SRK-86-079, dated November 17, 1986, was discussed with the QA
Manager. The QA Manager stated that the memorandum was issued to provide
a formalized means for QC inspectors to request information.or
clarification of field conditions but not as a substitute for NCR's.
Subsequent to the discussion, the QA Manager issued a revision to the
memorandum on January 14, 1987 stating that the use of memos "should only
be used when it appears that other mechanisms (NCR, ODR, etc.) are not
appropriate-for the condition.
The NRC inspector interviewed eight QC inspectors and the QC Supervisor
to determine if QC personnel clearly understood the licensee's NCR and
ODR policy and that NCR's or ODR's were written when required. The
sample included day and night shift QC inspectors, SMUD and U.S. Testing
QC inspectors and new (0 to 6 months) and old (6 months to 15 years)
Rancho Seco employee's. All of the QC inspectors interviewed-indicated
that they would write an NCR or ODR if one was required by applicable
procedures. None of the QC inspectors interviewed felt that there was
pressure not to write NCR's. However, during the interviews, the NRC
inspector identified the following concerns:
, -(a) Site specific QA/QC training for QC inspectors consists of
individual reading of the QA/QC procedures and a subsequent written
. exam. No formalized classroom site specific QA/QC training is
given. The QA/QC inspectors consequently have little opportunity to
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t !be fully' appraised of the procedural requirements, ask questions ~ and
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learn specific-details of the procedures, prior to~ performing actual
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-field inspections. ' Two inspectors' interviewed.had . limited knowledge
of site ~ specific requirements. .Pendi,ng furtherl inspection ofLthis-
E subject to determine significance,this was identifiedLas' Inspector
c Follow-up Item 87-06-10, "QC-Inspector Training.
(b) Two ofithe QC inspectors interviewed each identified-separate
nonconformance: reports (5703 and 6140) that had been written but had
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:been cancelled. The NCRs'and their cancellation were inspected by' ,
., the NRC-inspector. lNCR 5703 records indicate that the NCR was
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' voided, in accordance with QAP;17, subsequent to reasonable. .
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technical review by the licensee. .However NCR 6140 appeared to have-
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been voided without.. adequate review by the' licensee.
E Construction Inspection Data Report.(CIDR) Specification M17.08
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requires the inspection of the setscrews and pins of the governor
drive coupling lfor.the new TDI Emergency Diesel Generators. The
inspection required verification of- proper' installation by visually-
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determining that the " setscrews in place with locktite (i.e. they
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appear tight)!and are spot staked,"3and " pins installed'and' appear
. tight". xThe QC inspector assigned to perform the inspection'after-
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the 100 hr run of the 'dieselsi determined on December 9,1986 that
the " Governor drive coupling' set screws _were not; installed'using
Locktite, and were not staked in place", and initiated NRC-56140.
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On December.19, 1986, 10 days"after the first inspection, a
, different QC inspector' performed the subject inspection and
-determined the installationito be satisfactory. On January 10, 1987
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the QA Quality Engineer-completed the-CIDR with a note stating "Upon
re-assembly after inspection during phase "B", the setscrews were
installed w/Locktite 242 and spot staked. NCR #S-6140 should not
have been written.(per QAP #17) and will be voided." On January 21,
1987, NCR S-6140 was voided by the Engineering Review Board. On
February 7, 1987, subsequent to inquiry by the NRC. inspector
regarding NCR S-6140, the QA Quality Engineer cancelled the CIDR
note and replaced it with the following note "Further investigation
confirms that the set-screw (subject of NCR #S6140) was in place
"as-found". It was found to be tight.and spot-staked. NCR #S-6140
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was voided by the ERB on 1/21/87." The licensee was unable to
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obtain any documentation that work was performed to'Locktite and
stake the subject setscrews during'the NRC inspection.
10~CFR 50 Appendix'"B" Criterion XVI requires that " Measures shall
be established to assure that conditions adverse to quality, such as
failures, malfunctions, deficiencies, deviations, defective material
and equipment, and>nonconformances are promptly identified and
corrected. In the case of significant conditions adverse to-
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quality, the measures shall: assure that the cause of the condition
is determined.and corrective action taken to preclude repetition.
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- The identification of the significant condition adverse to quality,
the cause of the condition, and the corrective action taken shall be
documented and reported to appropriate levels of management.
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Contrary to the above, licensee measures, QA Procedure 17 Rev. 4,
Nonconforming Material Control, to assure that conditions adverse to
quality are promptly corrected are inadequate that QAP 17 paragraph
13.1 allows NCRs to be voided with Engineering Review Board approval
but does not provide a criterion for approval of voiding NCR's. In
addition, licensee measures were inadequate in that NCRs-6140 was
voided by the Engineering Review Board without determination of the
without determination of a need for
validity of the original
further corrective report,thout
action and wi documented corrective action
for the originally reported condition. Furthermore, the-licensee
measures were inadequate in that no determination was made to
determine the cause of the originally-(NCRs-6140) reported condition
regarding the newly installed emergency diesel generators or action
to preclude repetition for potential similar as-found nonconforming
conditions on the new emergency diesel generators.
The above noted condition was identified as violation 87-06-04,
" Inadequate NCR S-6140 Corrective Action." The licensee's QA
Manager disagreed that the condition was a violation because the
second inspector verified the installation to be acceptable 10 days
after the NCR was originated and the installation was again verified
to be acceptable on February 12, 1987. The NRC inspector noted the
QA Manager's dissenting opinion and informed the QA Manager that the
Violation still appeared to be appropriate in that NCR S-6140 was
cancelled without a reasonable attempt to reconcile the
unsatisfactory results of the first inspection and the satisfactory
results of the second inspection performed 10 days after the
original as-found condition was reported. Furthermore, no licensee
consideration was evident for the need for further 10 CFR 50
Appendix "B" Criterion XVI corrective action for potential similar
"as-found" deficient conditions for the newly installed TDI
Emergency Diesel Generators.
4. Staff Position
This allegation was not substantiated. The inspection verified that
there is reasonable assurance that nonconforming conditions are
being identified by QC inspectors. A violation regarding voiding of
NCR's was identified. Corrective action for Violation 87-06-04 will
be followed by the routine inspection program.
5. Action Required
Verify licensee corrective action for violation 87-06-04.
11. Receipt, Storage, and Handling of Equipment and Materials Program,
and Review of Procurement
The inspector reviewed the licensee's implementation of the QA program
and the material program relating to the control of receipt, storage, and
handling of equipment and materials for safety related equipment.
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-The inspector reviewed a sample of six receipt inspections and found that
the inspections appeared.to~have been generally conducted in accordance
with. Quality Assurance Procedure (QAP) 10, Receiving Inspection.
However, the following errors,were found: The review of two Receipt
Inspection Data Reports (RIDRs) for; Exxon Nebula EP0 grease, (Stock
Number 101861), revealed the storage locations identified on the RIDRs
were different for identical grease: one RIDR stated that the grease
should be stored in a level "B" storage area as defined by ANSI N45.2.2 -
1972, while the'other RIDR stated the same type grease should be stored
in a level "C" storage area. The inspector found various containers of
grease in a cargo storage container on site. The container only met the
ANSI requirements for level "C" storage. The licensee had not defined
its storage levels in a procedure other than the notation on the RIDRs.
The licensee committed to clarify this apparent storage level discrepancy
promptly.
The inspector reviewed the licensee's warehousing procedure,
AP 605, Rev. 11, " General Warehousing." The inspector's noted that the
following requirements existed: 10 CFR 50, Appendix B, Criterion V,
states, in part, " Activities affecting quality shall be prescribed by
documented instructions, procedures, or drawings, of a type appropriate
to the circumstances and shall be accomplished in accordance with these
instructions, procedures..."
Quality Assurance Program Policy Section V, Rev. O, " Instructions,
Procedures and Drawings", states, -in part:
"3.0 Policy
Activities affecting the operational safety or quality
performance at Rancho Seco shall be prescribed by and
implemented in accordance with documented instructions,
procedures, and drawings for the operational life of the
plant.
AP 605, Rev.11 requires, in part:
"3.4.2.6 All shelf life items and EQ items that have a shelf life
are sorted such that the oldest items are in front or_on top
and the newest in back or on the bottom.
3.4.3.4.1 A manual or computerized tickler file system will be set
up to ensure that specific storage control instructions... are
followed.
3.4.3.4.5.7 The person responsible for reviewing the tickler will
also be responsible for ensuring that the stock which has
reached its expiration date_is removed from the shelf on a
monthly basis..."
The inspector found that on February 24, 1987, no system had been
implemented'in accordance with AP 605 that sorted the shelf life items
such that the oldest items were.in front or on tos. A manual or
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computerized tickler file system was not implemented to control special
storage requirements; for example, the shelf life of. safety related
rubber components. No licensee personnel had reviewed or removed expired
shelf life items on a monthly basis.
The inspector found examples of expired shelf life 0-rings in the
A warehouse in their storage location and not removed. This included
0-rings intended for Class 1 equipment as in the case of Stock
Numbers 038318 and 113082.
This is an apparent violation of the quality assurance requirements for
control of materials (Violation 87-06-05).
The inspector also noted that 10 CFR 50, App. B Criterion VIII, states,
in part: " Measures shall be established for the identification and
control of materials, parts,'and components, including partially N
fabricated assemblies. These measures shall assure that identification
of the item is maintained by... part number, serial number... on the item
Quality Assurance Program Policy Section VIII, Rev. O, " Identification ,
and Control of Materials, Parts and Components" requires, in part
"3.0 Policy s
Appropriate procedural methods shall be prescribed and
implemented to assure that materials, spare parts or components
are properly identified and controlled to preclude the use of
incorrect or nonconforming items..."
However, the inspector observed that various safety related items in
stock in the Warehouses A and B were not identifiable. On February 24,
1987 no marking or tags with identification numbers were provided on the
parts.
This is an apparent violation of the noted requirements for
identification of safety related spare parts. Violation 87-06-06
The inspector also observed that:
A. Members of the warehouse staff, when asked, could not provide the
inspector a controlled copy of AP 605. Nor did a new employee who
was responsible for issuing parts know of the existence of AP 605.
The licensee later informed the inspector that a controlled copy of
AP 605 was available in Warehouse A.
B. Various techniques were used to apply the acceptance tag onto the
0-rings. This included the wrapping of a metal wire with a tag
connected to the wire around the 0-ring. This method of tagging
appreared to provide the potential for the cutting of the 0-ring
material by the wire.
C. The inspector reviewed the licensee's QA audit number 0-759
performed on 11/5-8/85 and 11/12/85, concerning activities performed
on all on-site warehouses (MSRC No. 22 Audit). The audit findings,
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in part, included: ~ Item 18, a finding'.that "... no special storage
requirements, shelf life, storage maintenance tickler file exists".
The-item was addressed by:t "This item will be placed on the QA
followup list (next MSRC No. 22 Audit)". The licensee would have
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next performed the MSRC.No. 22 Audit during of 11/86,~but due to a
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memo dated-November 17, 1986, the: licensee postponed the
investigation of this item and others to the nexttGeneral
Warehousing Audit _1.n late 1987. As discussed in this report, the
licensee to date, still does not have a tickler file as discussed in
AP.605. Therefore, the inspector is concerned by the method of the
quality assurance's verification.of corrective action to audit
findings.
D. -The inspector reviewed the contents of the cargo container where
.
.various spare parts for motor operator valves were stored. The
cargo container appeared to be appropriate only as a level "C"
storage by ANSI N45.2.2 - 1972. However, the inspector found
safety-related gaskets, 0-rings, and other parts which are
recommended by ANSI 45.2.2 to be stored in' level "B" storage.
These concerns will be followed as part of future inspections of the
licensee's response to the two violations in this subject area in this
report.
12. Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are-acceptable or may involve violations or
deviations. One new unresolved item identified during this inspection
is discussed on page 29.
13. . Exit Meeting
The inspectors met with licensee representatives =(noted in Paragraph 1)
at various times during the report period and fonnally on March 12, 1987.
The scope and findings of.the inspection activities described in this
report were summarized at the meeting. Licensee representatives
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acknowledged the inspector's findings and violations identified.
}}