IR 05000106/2002016
| ML17083B534 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon, 05000106  |
| Issue date: | 03/20/1985 |
| From: | Dodds R, Mendonca M, Mendoncaj M, Padovan M, Polich T, Ross T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17083B533 | List: |
| References | |
| 50-275-85-01, 50-275-85-1, 50-323-85-03, 50-323-85-3, NUDOCS 8504110406 | |
| Download: ML17083B534 (38) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos:
50-275/85-01 and 50-323/85-03 Docket Nos:
50-275 and 50-323 License No:
DPR-80 Construction Permit No.:
CPPR-69 Licensee:
Pacific Gas and Electric Company 77 Beale Street, Room 1451 San Francisco, California 94106 Facility Name: Diablo Canyon Units l and
Inspection at:
Diablo Canyon Site, San Iuis Obispo County, California Inspection conducted:
January 6 - February 16, 1985 Inspectors:
M. M. Mendonca, Sr.
Re ident Inspector ate Signed M. L. Pado an, esiden Inspector at Signed T,
M. Ross, es dent Insp tor D te igned T. J. Poli
, Resident Ins ector D
e gned Approved by:
R. T. Dodds, ief, Reactor Projects Section
Summary:
Ins ection from Janua
1985 throu h Februa
1985 Re ort Nos. 50-275 85-01 and 50-323 85-03 D
e S'gned Areas Ins ected:
Routine inspection of plant operations, conditions, and events; preoperational test program; startup test program; independent inspection; and followup of open items, allegations, and LERs.
This inspection effort required 245 inspector-hours for Unit 1, and 201 inspector hours for Unit 2 by four resident inspectors.
850411040b 850322 PDR ADQCK 05000275
DETAILS Persons Contacted
+R.
C. Thornberry; Plant Manager-R. Patterson, Assistant Plant Manager/Superintendent J.
M. Gisclon, Assistant Plant Manager for Technical Services
- W. B. Kaefer, Assistant Plant Manager for Support Services
"T.
W. Rapp, OSRG Chairman
""W. A. Wogsland, Technical Assistant to NPO Manager
- D. A. Taggert, Supervisor of Quality Assurance
-"C. L. Eldridge, Quality Control Manager
>R.
G. Todaro, Security Supervisor D. B. Miklush, Supervisor of Maintenance J,
A; Sexton, Supervisor of Operations J.
V. Boots, Supervisor of Chemistry and Radiation Protection
"-W. B. McLane, Material and Project Coordination Manager
~L. F.
Womack, Engineering Manager W.
G. Crockett, Instrumentation and Control Manager E. T. Murphy, Regulatory Compliance Supervisor T. J. Martin, Training Manager
"-W.
G. O'ara, Sr. Chemistry and Radiation Protection Engineer
"-R. P.
Powers, Sr. Chemistry and Radiation Portection Engineer Wt The inspectors interviewed several other licensee employees including shift supervisors, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, quality assurance personnel and general construction personnel.
"-Denotes those attending the exit interview.
0 erational Safet Verification During the inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities were conducted on a
daily, weekly or monthly basis.
On a daily basis, the inspectors observed control room activities to verify compliance with selected limiting conditions for operation as prescribed in the, facility Technical Specifications (TS)
~
Logs, instrumentation.recorder traces, and other operational records were examineL,to'btain information on plant conditions, trends, and coaipliance'ith, regulations.
Shift turnovers were observed on a sample basis<Fo'erify that all pertinent information on plant status was relayed.
During each week, the inspectors toured the accessible areas of the facility to observe the following:
(1)
General plant and equipment conditions.
(2)
Surveillance and 'maintenance activities.
(3)
Fire hazards and fire fighting equipmen (4)
Radiation protection controls.
(5) 'onduct of selected activities for compliance with the licensee's administrative controls and approved procedures.
(6)
Interiors of electrical and control panels.
(7)
Implementation of selected portions of the licensee's physical security plan.
(8)
Plant housekeeping and cleanliness.
(9)
Operability of selected Engineered Safety Features (ESF)
systems by performing comprehensive walkdowns of the system's components.
The inspectors'alked with operators in the control room, and other plant personnel.
The discussions centered on pertinent topics of general plant conditions, procedures, security, training, and other aspects of the involved work activities.
No violations or deviations were identified.
En ineered Safet Features (ESF) Actuation of Control Room Ventilation Inadvertant ESF actuation of the Control Room Ventilation System (CRVS)
occurred twice, on February 4th and 5th, due to Start-up (S/U) testing activities associated with Unit 2 Solid State Protection System (SSPS).
The common unit CRVS was automatically transferred from Mode 1 (normal operation)
to Mode 4 (pressurization),
during steps to restore from S/U test procedure 35,1 (Response Time Testing) for Unit 2.
Unit 1 was in operational Mode 1 (>5/ reactor power) during both events.
Resident inspectors were in the control room during course of the first event.
Licensee Event Report (LER) 84-19 issued on July 24, 1984 describes a
previous occurrence of a similar CRVS Mode 4 actuation event.
As corrective action for LER 84-19, to prevent future inadvertant CRVS mode shifts due to Unit 2 SSPS testing, circuit leads were lifted to isolate the specific CRVS initiating relay.
Approximately six months later S/U personnel reconnected these leads prematurely while performing the final steps of S/U TP 35.1.
This action allowed both CRVS mode shift events to occur unexpectedly when the restored ESF actuation circuit responded, as designed, to phase
"A" test signals from the Unit 2 SSPS.
The lifted circuit leads had been logged and identified with information tags, as required by administrative procedures.
However, the written information only referenced S/U TP 35.1 as the controlling document and did not clearly describe the original intent of the corrective action for LER 84-19.
These and other lifted leads were expected to isolate all common ESF equipment, from undesired actuations by the Unit 2 SSPS, until fuel load.
Corrective action resulting from recent events, has further clarified these instructions and will now require direct Shift Foreman (SFM) approval prior to restoring the lifted lead These events were promptly reported to the NRC Operations Officer and will be documented in a LER.
Resident inspectors reviewed the circumstances in detail and attended the responsible Technical Review Group (TRG) meeting.
No violations or deviations were identified.
4.
Startu Test Re ort The licensee's Startup Test Report for the period from fuel load to completion of the special low power tests was enclosed with PG&E letter No. DCL-85-031 dated January 29, 1985, The report accurately described this startup test program, that was observed and documented in previous inspection reports.
No violations or deviations were identified.
5.
Maintenance The inspectors observed portions of, and reviewed records on, maintenance activities to assure compliance to approved procedures, technical specifications, and appropriate industry codes and standards.
a
~
Corrective Maintenance of Pressure Indicator Pressure indicator 942 for the accumulators filland test line was removed from service, dismantled, and repaired.
Selected clearance points and associated verifications were observed by the inspectors, as well as, satisfaction of radiation protection and cleanliness requirements.
The pressure indicator was then calibrated and returned to service.
b.
Reactor Coolant Pum (RCP) Under Volta e (UV) Rela s
Selected aspects of an electrical maintenance activity to replace RCP UV relays were observed by the inspector.
All 12 KV Bus D and E
RCP UV relay devices type SV (mechanical)
were replaced with solid-state type S-SVT devices for improved reliability per Design Change Notice (DCN) DC2-E-E1786.
Portions of electrical work observed by the inspector included:
1) determination and reconnection of UV relays, 2) jumper installation and removal, and, 3)'lectrical cabinet alterations.
Furthermore, the inspector reviewed the approved Shop Work Follower (SWF) EM-1-85-007, authorized Zlearances, and maintenance procedure (MP) E-5,33
"Routine Preventive Maintenance of Type SSV-T Relays."
To verify operability of RCP UV relays and restore them to normal service, ISC technicians performed post maintenance surveillance testing in accordance with Surveillance Test Procedure I-9A and B,
"Functional Test of 12 KV Channels" and "Removal from Service, Calib and Reinstatement for 12 KV UV Protection and Safequards Functions,"
respectivel c
~
SG Blowdown Isolation Valve Gasket Re lacement As excessive body-to-bonnet leak was identified and documented in Nuclear Plant Problem Report (NPPR) for SG 1-4 blowdown isolation valve MS-1-FCV-763.
SWF jj MM-1-85-021 prescribed the procedure for the-step-by-step performance to replace gaskets in this valve.
The inspector reviewed all applicable documentation associated with this mechanical maintenance evolution, which included the NPPR, SWF, Quality Control (QC) instruction No. 85-0027, and authorized clearance 4-251-85.
Furthermore, the inspector cross checked gaskets identified by material request form No. 8948 against those recommended in the manufacturor's manual for. "Fischer Control Valves."
In accordance with the SWF and QC instruction for valve re-assembly following gasket replacement, final torquing was accomplished in specified increments to a maximum allowed value.
Radiological controls were established by special work permit (SWP)85-105 and maintained accordingly.
Loo Seal Fill of Pressurizer Safet Valve As an attempt to resolve continued problems with elevated temperature indications in downstream discharge piping of pressurizer safety valve 8010B, the upstream loop seal was refilled in accordance with SWF ff MM-1-85-028.
As a consequence of this evolution, flange gasket and bolting associated with the loop seal drain header vent line were also replaced.
The inspector monitored work activities of mechanical maintenance, radiolgical control, and QC personnel.
Appropriate radiological contamination controls were in effect at all times.
e.
The inspector reviewed in detail and observed implementation of the following:
1)
SWF /jMM-1-85-028, prescribed step-by-step work activity conduct, 2) clearance 7-313-85, provided authorization and clearance points, and 3), QC instruction /j85-0045, established hold points for QC inspection during SWF performance.
Fire Water S stem Valve Portions of. preventive maintenance on fire water system valve 18 was observed by the inspector.
The work was performed in accordance with.an approved SWF and QC hold points were observed.
The maintenance-entailed disassembly, inspection, lubrication, and reassembly of the valve.
Proper procedures wer'e followed for sealing shut the valve after maintenance.
f.
Containment Pur e Exhaust Fan Portions of corrective maintenance on the subject, Unit 2 fan, E-3, were observed by the inspector.
Selected clearances and work instructions were followed.
QC hold points and testing requirements were specifie No violations or deviations were identified.
Surveillance By direct observation and record review of licensee surveillance testing, the inspectors verified compliance with Technical Specification (TS)
requirements and implementing plant procedures.
a
~
Functional Test of Power 0 crated Relief Valve (PORV) Over ressure Protection Channel b.
The inspectors observed portions of a functional test on the overpressure protection channel for PORV 456.
This test verified PORV overpressure protection pressure and temperature setpoints and instrumentation logic.
Surveillance was performed in accordance with licensee reviewed and approved Surveillance Test Procedure (STP) I-69A in order to satisfy TS monthly analog channel operational test requirements.
The tested channel was removed from service and later returned to service in accordance with the STP.
Communication with the control room operators was maintained during testing to assure operator awareness on TS operability requirements.
Data sheets were accurately and neatly kept, and subsequently reviewed by an Instrumentation and Controls foreman.
Functional Test of the Post-Accident Sam le Room Area Monitor c ~
The inspectors observed a functional test of radiation monitor (RM)
48.
The test was conducted in accordance with STP I-18AA1.
This STP verified, (1) appropriate channel response to a check source, (2) high and alert alarm indications functioned, (3) the high and alert setpoints were correct, and (4) the failure alarm was operable.
Communications with the control room was maintained for verification and acknowledgement of alarm indications.
Improvements that should be made in the STP and associated data sheet were noted to an appropriate level of management by the auxiliary operator.
The test acceptably verified RM operability.
Stroke Time Testin of Accumulator Dischar e Valve The subject testing was conducted in accordance with STP V-3L4.
The technician was knowledgable of the testing alignments, precautions and acceptance criteria.
Contact with the control room was maintained to assure compliance with TS requirements.
Test equipment was calibrated.
The test results were reviewed by control operator and shift foreman, and the system was acceptably realigned.
d.
Letdown Orifice Isolation Valves 0 erabilit Post maintenance operability of letdown orifice isolation valves 8149 A, B, and C were tested in accordance with STP V-2Y and STP V-3K7A.
The inspector observed an operator exercise all three valves to verify operability, position indication, and protective interlock features as required by STP V-2Y.
Stroke time testing of each valve in accordance with STP V-3K7A (to verify compliance with
TS 3.6.3 Phase
"A" Containment Isolation Closure requirements)
was also performed by operators and observed by the inspector.
During valve exercises and stroke time testing, an operator recorded all applicable data and checked off each performed step.
Both completed STPs were subsequently reviewed and approved by a shift foreman.
Operators conducted both STP's in accordance with required instructions.
Recorded data was determined to meet all established acceptance criteria for each valve.
The shift foreman subsequently reviewed STP V-2Y and V-3K7A as satisfactorily complete.
e.
RCS Tem erature Instrumentation The inspector observed portions of calibration procedure STP I-5B2 for Tavg, Delta-T, overtemperature Delta-T, and overpower Delta-T channels.
The technicians removed the system from service in accordance with the approved procedure and in compliance with TS requirements.
These temperature channels were acceptably calibrated using the required instrumentation.
RCS Leaka e Determination The inspector observed an auxiliary operator in the performance of-STP R-10B.
The calculation of unidentified leakage by inventory balance of containment building sumps was observed.
Additionally, verification of containment building temperature monitors was observed.
Problems with the data review were identified to
'appropriate levels of licensee management by the inspector and will be followed under normal inspection effort.
Compliance to TS unidentified leakage rate requirements were acceptably verified.
No violations or deviations were identified.
7.
Routine Ins ection a
~
Document Control An inspection of the licensee's Document Control Program was conducted during the inspection period.
The inspection included both a review of procedures governing document control and a
verification of the implementation of those procedures.
The-licensee has administrative procedures for the control of drawings, procedures, design changes and other records required to be'aintained by American National Standards Institute (ANSI)
N18.7-1976 and ANSI N45.2.9-1984.
Master indexes are maintained for drawings, procedures, technical manuals and other documents which indicate the current revision or change.
The inspector verified by a representative sample that the control room copy and other copies of procedures and drawings were of the same revision as the site master copie b.
Units 1 and 2 0 erator Re uglification Trainin Pro ram Within three months following issuance of a facility operating li'cense, implementation of a requalification training program for, Reactor Operators (RO) and Senior Reactor Operators (SRO) is mandated by Federal Regulations (10 CFR 50.54).
As prescribed in 10 CFR 50.34, this program must be established in accordance with a descriptive plan contained within the Final Safety Analysis Report (FSAR) which at a minimum, meets all the criteria specified by Appendix A of 10 CFR 55.
Furthermore, NUREG-0737 ("Clarification of TMI Action Plan Requirements" ) requires all power reactor licensee's to revise their requalification training program to be consistent with supplemental guidance provided in the NRC letter from H. Denton dated March 28, 1980.
In addition to Appendix A of 10 CFR 55 and the letter in NUREG-0737, Technical Specification (TS) 6.4 directs the facility staff to comply with recommendations in ANSI N18.1-1971.
It was the purpose of this routine inspection to verify an on-site requalification training program has been established and effectively conducted in accordance with approved procedures based upon aforementioned regulations, requirements, and recommendations.
Within the scope of the inspection process, the following documents, procedures, records, etc.
were reviewed in detail:
TS 6.4, "Training" Appendix A of 10 CFR 55, "Requalification Programs for Licensed Operators..."
NUREG-0737 Section I.A.2.1 (NRC letter from H. Denton dated 3/28/80),
"Immediate Upgrading of RO and SRO Training and Qualifications" NUREG-0737 Section I.A.2.3, "Administration of Training Programs"
,Updated FSAR Section 13,2.2,
"Licensed Operator Retraining Program" Regulatory Guide 1.8, "Personnel Selection and Training" ANSI St'andard N18.1-1971 Section 5.5, "Retraining and Replacement Training" ANSI/ANS Standard 3.1-1978 Section 5.5, -"Operator Retraining and Replacement Retraining" Nuclear Plant Administrative Procedure (NPAP) B-101 Revision 3,
"NRC Licensed Operator Retraining Program" NPAP B-101S1 Revision 0 and 1, "Supplement 1 to AP B-101 NRC Licensed Operator Requalification Progra NPAP B-151 Revision 0, "Operations Department Training Records".
General:. Office equality Assurance Program Audits dated 10/24/83 and 9/14/83, "Training and qualification of Plant Personnel".
Nuclear Plant Operation (NPO) Audits dated 8/10/83 and 9/18/83,
"Training Review".
1983 and 1984 Schedules for Simulator and Seminar Training.
Personnel records containing Annual Exams, Weekly quizzes, Oral Exams, Control Manipulations, Attendance, Exemptions, and Performance Evaluations.
Simulator Training records documenting attendance, performance of required manipulations and evolutions, and instructor evaluations.
The inspector attended a preplanned classroom lecture on emergency procedures and observed several simulator training sessions.
Several licensed ROs and SROs were interviewed by the inspector to, verify content of personnel records and participation in the requalification program.
Futhermore, during the course of inspection, the following training personnel were interviewed:
Operations Training Supervisor, Requalification Training Coordinator, a licensed contract instructor, Simulator Training Supervisor, an Assistant Operator Training Coordinator, and the personnel records clerk.
After completing the aforementioned inspection activities, the inspector has determined:
1)
The licensee has established a requalification training program which satisfactorily meets NRC regulations and requirements, and industry standard's recommendations.
2)
Implementation of requalification training has been effectively conducted in accordance with approved procedures based upon the descriptive plan in chapter 13 of the FSAR.
Several areas were perceived by the inspector as requiring some improvement in order for them to fully meet the guidance established in con<rollTng procedures.
Those subject areas which were discussed with the licensee's training staff included:
records control, use of exemption sheets, inconsistent documentation of on-shift manipulations/evolutions, and clarification of the responsibility for-providing industry operational experience to the training department as defined by TS 6.4.
c.
Test and Ex eriment Control The inspector examined the licensee's program for the control of tests and experiments involving safety related structures, systems,
'
'
and components.
.The inspector reviewed the requirements and commitments of 10 CFR 50,59, ANSI N18.7-1976, FSAR Chapter 17, and TS.,chapter 6 as they relate to test control. It appears the licensee presently meets the minimum program requirements.
The licensee has recently issued a procedure which covers not only changes to test and experiments, but also changes to the facility and to procedures that require a
CFR 50.59 Safety Evaluation.
This procedure when fully implemented will provide added assurance the licensee's commitments and regulatory requirements are met.
The licensee's General Office issued Nuclear Power Generation Department Procedure NPG-5.14,
"10 CFR 50.59 Safety Evaluation Guidelines,"
approved ll/26/84.
This procedure establishes the licensee's position on the types of proposed changes, tests or experiments that require a written Safety Evaluation as-specified in
CFR 50.59, and guidelines for identifying those changes requiring such a Safety Evaluation.
The Plant Staff Review Committee (PSRC)
has agreed to implement this procedure on-site within a month.
The PSRC's implementation of this procedure will be followed as a part of the routine inspection program.
No violations or deviations were identified.
8.
Alle ation Followu Task:
Alle ation or Concern No.
1101 ATS No:
RV-84-A-064 a
~
Characterization Falsification of records - American Society for Nondestructive Testing (ASNT) Level II Magnetic Particle Test (HT) inspector restrictions were waived by a backdated signature of a Pullman I,evel III MT inspector who was no longer on site.
b.
Im lied Si nificance to Plant Desi n Construction or 0 eration The qualifications of Level II MT inspectors were compromised by falsification of records.
C.
Assessment of Safet Si nificance The.-records of the Level II MT inspectors in question were examined as well as the records of other inspectors who were certified during the same time frame.
The staff verified each of the Level II MT inspectors met all ASNT requirements including experience as specified in ASNT Recommended Practice No. SNT-TC-1A at the time of certification.
The staff identified letters in the training records which placed interpretation restrictions on several Ievel II MT inspectors.
These restrictions were placed on the inspectors several months
~
'
after initial qualification.
The staff interviewed the Field QA/QC Manager who placed these restrictions on the Level II inspectors and found that the restrictions were placed on the inspectors verbally at;,the time 'of qualification.
Furthermore, these interpretation restrictions required all results of examinations conducted by these I,evel II inspectors to be subject to a final interpretation by a Level III MT inspector.
This restriction remained in effect until the individual Level II inspector could interpret examination results to the satisfaction of the Field QA/QC Manager, who was a
qualified I,evel III MT inspector.
The restrictions were removed by verbal notification of the individual inspector and by placing a
letter removing the interpretation restrictions in the inspector's training record.
The several month lag time in placing the restriction letters in the inspectors training records was due to the hospitalization of the Field QA/QC Manager.
The staff also reviewed an audit of training records conducted by a QA internal auditor which identified two inspectors who did not have letters removing these restrictions in their training records.
The QA internal auditor who had identified these training record discrepancies closed the audit finding by allowing the new Field QA/QC Manager to forward backdated letters to the former Field QA/QC Manager, who had placed the restrictions on the two inspectors.
The former Manager who had'been transferred to another job site signed these letters which were backdated to when he was last on site at Diablo Canyon.
These backdated letters were then placed in the training record of-the two Level II inspectors with no indication that the letters were backdated or the reason for backdating.
The staff reinterviewed the former Field QA/QC-Manager and found that the two Level II MT inspectors had both been given verbal notification that they could resume full interpretation responsibility as they had demonstrated satisfactory interpretation ability to the Manager.
The former Field QA/QC Manager admitted the failure to ensure a letter documenting the restrictions were removed was an administrative oversite on his part.
The Manager interpreted his signing of backdated letters as an acceptable method of resolving an administrative oversite on his part to th'e satisfaction of the QA auditor.
d.
Staff. Position The;-'staff finds that the inspectors met all code and regulatory requirznents and therefore, finds no violations of regulations.
The issue here concerns the administration of requirements placed over and above those" required by the Code.
The staff finds the backdating of records an unacceptable practice in any circumstance let alone as a resolution to an audit finding and believes the licensee should take corrective action to preclude reoccurrence.
, However, even though an unacceptable practice was used, it had no safety significance.
The qualifications of the Level II MT inspectors was not compromised as a result of the backdated letters as they had both met all requirements of SNT-TC-lA prior to certification and prior to having the restrictions placed on them
and they did not start interpreting on their own until given permission to do so by the (}A/(}C Manager.
e.
Action Re uired No further action is required.
Task:
Alle ation or Concern No. 718-ATS No:
RV-84A-062 a
~
Characterization I'nstructional documents on weld symbology were not distributed as claimed by PG&E.
b
~
Im lied Si nificance to Plant Desi n Construction or 0 eration This GAP rebuttal implies PG&E has made false statements concerning distribution of weld symbol handouts and clarity of memos concerning weld symbology interpretation.
c
~
Assessment of Safet Si nificance As a result of investigation into allegations listed in SSER 21, the NRC staff requested Pacific Gas and Electric (PG&E) to provide further information on the use and interpretation of weld symbols.
In response, PG&E issued letter no. DCL-84-40 dated February 7,
1984.
Within this letter, under the subject heading of
"Communication of Information", statements were made that
"Communication on weld design and weld symbol use has taken several forms, including discussion sessions and written direction...this program will continue to assure proper communication of weld symbol use and weld design (Attachment 3)."
Attachment 3 of PG&E letter DCL-84-40 was enclosed as a copy of the weld symbol handout distributed at discussion (training) sessions.
Examples of written directions circulated to the field were also enclosed in PG&E letter DCL-84-40 as Attachment 4.
Under the topic heading
"PG&E Position 16" in attachment 5 of the March 23, 1984, Government Accountability Project (GAP) letter, the
, aforementioned PG&E response was refuted by GAP as follows "Further, as of March 16, 1984, the referenced Attachment 3 was not issued to the field, or at least to anywhere that I or anyone I know at Diablo Canyon has worked".
PG&E has subsequently addressed the GAP rebuttal in their letter DCL-84-243 dated June 29, 1984.
In this response, PG&E declared the GAP statement to be false and re-affirmed that distribution of Attachment 3 to DCL-84-40 had occurred during ",jobsite training programs regarding weld symbols which were conducted in May, June and July of 1983, in which 350 engineers and inspectors received training in weld symbols".
l
The inspector has reviewed all previously mentioned GAP and PG&E correspondence.
Conduct of training sessions and distribution of Attachment 3 type weld symbol handouts were verified by the inspector and documented in Inspection Report 50-275/84-36.
Handouts of this kind were distributed during training sessions, they were not circulated openly to the field.
Nor did PG&E claim that such was the case.
A corollary issue identified in GAP rebuttal of "PG&E Position 16",
also disputed the content accuracy of weld symbology clarification correspondence that were enclosed in Attachment 4 to PG&E letter DCL-84-40.
The inspector reviewed the subject memos which were titled dated October 10, 1983,
"Do's and Don'ts for Welding Symbols" and "Clarification of Pipe Support Weld Symbols" dated October 10, 1983.
Content adequacy of these weld symbol clarification memos was determined to be acceptable by the inspector considering their intent and context of use, as the clarification memorandum dealt with symbols applicable to existing documents and the "Do's and Don'ts" related to future work.
The NRC staff formally addressed this issue in Allegation 720 and was closed (as complete)
by Region V Allegation Board meeting of October 24, 1984.
The fundamental technical issue concerning adverse impact on safety-related construction activities from weld symbol misinterpretation was addressed by the staff in SSER-22 (allegation No. 126).
Other allegations (Nos.
719, 1116, and 1118) concerning
"ambiguous weld symbology" and "inadequate weld symbol training" were also resolved by the staff in Inspection Report 50-275/84-36.
The inspector could discover no evidence, nor discern any reason, to support the GAP rebuttal.
Applicable responses in PG&E letters DCL-84-40 (dated February 7,
1984)
and DCL-84-243 (dated June 29, 1984) appear to be consistent and acceptable.
d.
Staff Position The inspector concludes this allegation is unsubstantiated, and any technical significance has been previously addressed by the staff in documents referenced above.
e.
No further action required.
Task:
Alle ation or Concern Nos.
1649 and 1654 ATS No:
RV-84-A-122 a
~
Characterization Promatec installation of seals in crane wall (inside containment)
and auxiliary building penetrations is not satisfactory since the seals are being "faced off" with material which does not have traceabilit I lied Si nificance to Plant Desi n
Construction or eration The subject penetration seals perform one or all of the following functions:
l) radiation shielding, 2) hydrostatic/air sealing and 3) fire barrier sealing.
Unsatisfactory penetration shielding could result in increased personnel radiation exposure or unanticipated damage to safety-related equipment (as a result of fires or steam line breaks).
Assessment of Safet Si nificance The concerned individual identified specific locations to the licensee where face-off material, which did not have material traceability, was applied.
As described in (}uality Concern Summary Report (gCSR)-110, the licensee then documented these discrepancies in Minor Variation Report (MVR) C-1471, Rev.
1, since Promatec had failed to properly document the use of untraceable materials.
PGSZ Engineering and Promatec subsequently evaluated this situation and determined that the depth of the shield material could typically be varied by plus or minus two inches overall.
Accordingly, Promatec issued Document Change Request 0071039 to allow a two inch underfill tolerance on penetrations (mechanical penetrations with depth greater than eight inches; electrical penetrations with depth greater than twelve inches).
Thus, the face off material installed in the penetrations was determined to be unnecessary, and material traceability of that material was no longer required.
In order to prevent reoccurrence of this type, Promatec (}C and craft personnel were reindoctrinated in procedural requirements pertaining to documentation and traceability for facing-off penetrations.
Furthermore, additional actions were taken by the licensee to assure that untraceable material would not be used in penetration seals.
Staff Position The staff finds that the described discrepancy has been properly documented and dispositioned; and that although the licensee failed to assure that. face-off materials were traceable, there would be no adverse impact on safety if the face off material were not traceable.
A'chion. Re uired No further action is require Task:
Alle ation or Concern No.
1650 ATS NO:
RV-84"A-122 a.
Characterization High density lead elastomer (HDLE) material, used in penetration seals (see Allegation or Concern No. 1649),
had material expiration dates which were exceeded.
b.
Im lied Si nificance to Plant Desi n Construction or 0 eration Same as Allegation or Concern No.
1649.
c.
Assessment of Safet Si nificance In gCSR-110, the licensee verified that a shipment of outdated lead elastomer was received on-site, but had been placed on "hold" as indicated in Promatec Receiving Report Nos. 007/048, 007/051 and 007/052.
Promatec removed expired bags of elastomer from the HDLE barrels (which also contained dry lead particles)
and replaced them with new certified elastomer.
A check of several storage areas showed that material currently in use was not outdated.
d.
Staff Position As the outdated elastomer had already been properly identified and not used in penetration seals, no discrepancy was found to exist.
e.
Action Re uired No further action is required.
Task:
Alle ation or Concern No.
1651 ATS No:
RV-84-A-122 a.
Characterization Catalist and accelerator materials, used in penetration seals (see Allegation or Concern No. 1649),
were used onsite even though this material had passed its expiration date.
b.
I li'ed Si Kificance to Plant Desi n Construction or 0 eration C.
Same as Allegation or Concern No.
1649.
Assessment of Safet Si nificance The licensee's investigation revealed that Promatec did use expired cure accelerator materials in penetration seals, but these errors were documented on Nonconformance Reports 007/010 and 007/019 (see gCSR-110), prior to the concerned individual identifying the problem to the NRC.
Shortly thereafter, a sample of the cure accelerator
from the same lot was tested by the manufacturer and deemed acceptable, and the useful life was extended.
Accordingly, the cure accelerator material used in the penetrations was also acceptable.
In order to provide better control of seal material shelf life, Promatec issued Site Work Instruction PMT-007.001 "Field Monitoring and Control of Promatec Sealing Materials with Shelf Life Limitations."
The procedure requires monitoring of controlled materials which have shelf lives, and defines personnel responsibilities, material control methods and corrective action requirements.
d.
Staff Position The staff has determined that the licensee had already identified and initiated corrective actions to resolve the issues raised in this concern.
e.
No further action is required.
9.
0 en Items Followu a
~
Masonr Walls Interface with Safet -Related Com onents (TI-15-37 closed Units 1 6 2 The licensee has reviewed the analysis for masonry wall structural integrity in the Independent Design Verification Program and the Internal Review Program.
This review used more conservative computer codes than originally used for masonry wall design evaluation.
This revised analysis showed that additional support was required for selected masonry block walls.
NRR is continuing to follow the licensee's analysis required by License Condition 2.C.(10).
The inspector verified completion of selected portions of the additional support requirements.
This included verification of angle iron-base plate installation (through-bolting and anchoring)
to the building structures.
This closes the open temporary instruction.
No violations or deviations were identified.
b.
Ioss of Source Ran e Instruments (o en item 84-21001 closed)
The licensee has changed applicable STP's which should preclude future loss of source range instruments.
Furthermore, the generic issue of on-the-spot changes to procedures has been addressed and how such changes are incorporated into the text of the procedure, This item is closed.
10.
Isolations of Residual Heat Removal (RHR)
S stem Unit
On January 20, 1985 at approximately 11:00 p. m., the RHR system hot leg suction line was unexpectedly isolated.
Operators were alerted by the tl
RHR low flow alarm and immediately secured the RHR pumps.
Subsequent investigation by the operators determined an inadvertant isolation of the RHR system had occurred as a result of high RCS'pressure due to surveillance testing by Instrumentation and Control (ISC) personnel.
The RHR system was returned to service in about ll minutes after diagnosis and reinstatement of the instrumentation and controls system.
ISC personnel were performing surveillance test procedure I-68A to ascertain instrumentation logic functional response of the RCS wide range pressure channel.
This procedure required the technician to verify an open power supply breaker to the RHR hot leg isolation valve to preclude inadvertent isolations.
This valve would receive an automatic closure signal on high reactor coolant system pressure as a protection feature from inter-system loss of coolant accidents.
While performing verification of the breaker position, an IGC technician
.found the correct breaker, but mistakenly verified the position of an adjacent breaker with a caution tag.
Therefore, during the surveillance test, a high pressure signal was simulated which caused the associated RHR suction valve to close as designed.
Plant management conducted an Incident Review Board to critique the event.
In attendance were the responsible ISC technicians and Foreman,.
as well as, the IRC General Foreman and Manager, two Assistant Plant Managers (Plant Superintendent and Technical Services)
and the Compliance Supervisor.
The Incident Review Board concluded:
1)
a procedure change was desirable to describe the impact on plant conditions if the breaker position is not properly verified and that operator action may be needed to open 'the breaker, 2) lesson plans for operators will be reviewed to assure the RHR isolation function is discussed, and 3) improvements of independent verification activities should be considered.
The Incident Review Board decided that personnel error was the fundamental cause of the event.
All involved personnel were appropriately counseled concerning this problem and its ramifications.
Additionally, the IRC manager reinforced the importance of a verification process as department policy among all IRC personnel.
More specifically, ISC has a policy which requires independent verification for "removal from service" or
"reinstatement to service" during surveillances.
Procedural formalization of this policy should be included in the licensee's evaluation of the independent verification program.
On January 25, 1985 another RHR isolation along with a diesel generator start occurred'ue to an electrical power spike on the instrumentation bus tha<'.provides control power to the hot leg isolation valves.
This isolation. occurred during testing of power supply breakers and is not due to. equipment failure.
Corrective maintenance of power supply breakers acceptably addressed this event.
The licensee plans to report these events as licensee event reports.
No violations or deviations were identifie Prep erational Testin (Unit 2)
The inspectors observed selected portions of the preoperational test on fuel handling and transfer equipment.
The test verified function of
'andling tools, transfer carriage, and associated protective instrumentation (load cells and position limit switches).
The test was conducted in accordance with startup test procedure 34.1.
Test personnel were knowledgable of test conduct and administrative controls.
No violations or deviations were identified.
I,icense Event Re ort (LER) Follow-u (Unit 1)
Circumstances and corrective actions described in LERs, listed below, were examined.
The inspectors verified these LERs were reviewed by the licensee, and reported to the NRC within required time intervals.
The inspectors also ensured appropriate corrective actions were established and applicable events were accurately described.
Accordingly, the following LERs are considered closed.
LERs 85-01, 85-02, and 85-03 were discussed in Inspection Report No.
50-275/84-40.
No violations or deviations were identified.
Safet In ection At approximately 12:43 a.m.
on February 13, 1985, Diablo Canyon Power Plant Unit 1 experienced a safety injection and an associated reactor trip.
This event was caused by spurious protective channel signals of high differential pressure between steamlines while at 75 percent power for startup testing (incore flux mapping was in progress).
All plant equipment operated as designed.
Plant operators responded in accordance with emergency operating procedures and stabilized the plant in approximately 30 minutes.
The spurious safety injection signal occurred on actuation of two-out-of-three independent protective instrument channels.
The licensee's investigation of the spurious signal determined that the cause of the event was radio interference to the pressure transmitters from a security radio.
The licensee has precluded operation of radios in the vicinity.-of instrument transmitters.
The licensee had previously conducted'alkthrus with radios to determine potential points of interference, however, changeout of pressure transmitters has not considered this effect.
The licensee plans to fully evaluate the potential for radio interference prior to allowing future use of radios in the areas where there are instrument transmitters. 'n Unusual Event was declared at 12:50 a.m.,
and was terminated at 1:18 a.m.
The licensee made notifications of the event in accordance with emergency procedures.
The licensee plans to report this event as an LER.
No violations or deviations were identifie.
Exit Interview'n exit meeting was conducted with the licensee's representatives identified in paragraph 1.
The inspectiors summarized the scope and findings of the inspection as described in this repor