IR 05000361/1981034
| ML20040H532 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/02/1982 |
| From: | Bishop T, Eckhardt J, Wagner W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20040H529 | List: |
| References | |
| 50-361-81-34, 50-362-81-09, NUDOCS 8202180348 | |
| Download: ML20040H532 (10) | |
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U. S. NUCLEAR REGULATORY C0fHISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION V
50-361/81-34 R0 port No. 50-362/81-09 Docket No. 50-361, 50-362 License No. CPPR-97, -98 Safeguards Group Licensee: Southern California Edison Company 2244 Walnut Grove Avenue
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Rosemead, California 91770 Facility Name: San Onofre Units 2 and 3 Inspection at: Construction Site, San Diego County, California Inspection conducted: November 16-20 and December 7-11, 1981 Inspectors: A
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. Eckhardt, Reactor Inspector Bate Signed
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W gner, Rea r Inspector
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S b1 Approved by:
T. W. Bishop, Chief 1 Date Signed Reactor Projects Section 1
Summary:
Inspection on November 16-20 and December 7-11_, 1981 (Report Nos. 50-361/
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81-34 and 50-362/81-09)
Areas Inspected: Routine, unannounced inspection by regional based inspectors of construction activities involving licensee action on 50.55(e) items, IE Bulletins, and instrumentation and electrical cable quality records review.
The inspection involved 100 onsite inspection hours by two NRC inspectors.
Results: One item of noncompliance was identified in the area of removal of completed equipment without proper documentation (paragraph 7).
8202180348 820203 PDR ADOCK 05000361 G
PDR RV Fonn 219 (2)
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DETAILS 1.
Persons Contacted a.
Southern California Edison Company (SCE)
- J. M. Curran, Manager, Quality Assurance
+ D. E. Nunn, Project Manager
+*D. B. Schone, Site QA Supervisor
+*D. C. Stonecipher, Construction QA Supervisor
+ C. R. Horton, Startup QA Supervisor
+*N. M. Ferris, QA Er.gineer
+*W. Kirby, QA Engineer
- G. Vaslos, QA Engineer
+ J. E. Raniere, QA Engineer Trainee b.
Bechtel Power Corporation (Bechtel)
+*L. W. Hurst, Project QA Manager
+*J. W. Sheppard, Assistant Project QA Manager
+*J. H. McCarty, QC Manager
- D. W. Graf
- K. G. Hess
+ D. T. Lobree
+ D. W. Stohman
- Denotes those attending exit meeting of November 20, 1981.
+ Denotes those attending exit meeting of December 11, 1981.
2.
Licensee Action on Previous Inspection Find,i m The inspector examined the action taken by the licensee on the following items:
a.
(Closed) Followup Item (50-361/80-19/01):
Non Class lE instrument tubing connected to Class 1E instrument tubing.
It was 'etermined-that the non Class lE gressurizer and steam d
generator level instrument tubing was installed in accordance with the Bechtel QA program. meeting.10 CFR 50 Appendix B requirements.
Also, the inspector reviewed a letter.from Foxboro to SCE dated May 22, 1981 certifying that the Model E-130 Series level transmitters used in the non Class 1E portion were seismical_ly tested in accordance with IEEE 344-1971.
This item is closed.
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(Closed) Followup Item (50-361/81-21)
During IE inspection 81-21, the inspector discovered, when reviewing N-5 data package S2-1212-01, that radiographic reader sheet PBT-RT-10597 indicated that the weld was acceptable, however a piece of weld wire was located in the pipe. The wire was located near weld number CD on line S2-1212-ML-005 sheet 3 of the radiation monitoring system.
Although this is a non-safety related system the inspector inquired as to the present location of the wire, and what significance or potential affects the enclosed wire could have on the safe operation of the plant.
The licensee addressed this item in NRC Action Item Report (NRCAIR)
No. F-NRC-326. The inspector reviewed the Field Inspection Report (FIR) dated 9/26/81.
The FIR showed that radiographs were taken of the valve and pipe assembly. The radiographic reader sheets (RT-1306 and RT-13917) indicated that the wire was no longer in the system.
Physical examination of the system by the inspector revealed that the pipe size and configuration would prevent the wire from going beyond the area reported.
In addition, the non-safety related valve could easily be dismantled to remove the wire if it were in the system. The inspector is satisfied that the wire is not now in the system and has no further questions on this. item.
3.
Licensee Action on IE Bulletins The foll_owing IE Bulletins were reviewed by the inspector to determine the promptness and thoroughness of licensee actions _to correct or avoid those known or potential deficiencies:
a.
Bulletin 80-05, Vacuum Condition Resulting in Damage to Chemical Volume Control System (CVCS) Holdup Tanks Bechtel has performed a design review of all radioactive tanks i
subject to vacuum conditions. Three tanks were identified as requiring negative pressure relief protection; two Spent Resin l
Tanks (T-059 and T-060) and the Backflushable Filter Crud Tank (T-073). The design reviewed also showed that vacuum relief breather valves have been provided on the following seven tanks:
(1) Concentrated Miscellaneous Wastes Tank (T-061)
(2) Concentrated Miscellaneous Wastes Tank (T-062)
(3) Miscellaneous Wastes Tank (T-063)
(4) Chemical Waste Tank (T-064)
(5) Concentrated Boric Acid Storage Tank (T-069)
(6) Nuclear Condensate Tank (T-075)
(7) Nuclear Condensate Tank (T-076)
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-3-NRC. Action Item Report (NRCAIR F-NRC-204) was initiated to track and status actions associated with providing vacuum protection on the three tanks. The long term corrective action is to provide vacuum breakers and bypass lines on these tanks.
However, the vacuum breaker valves cannot be obtained prior ~ to fuel load, therefore, the licensee.has incorporated administration controls in the operating procedures for draining / fluid pumping of these tanks to preclude damage by vacuum conditions. The following procedure changes were written to reflect the Bechtel recommendations for precluding short-termitank damage during plant operation:
(1) Procedure Change Notice (PCN) to Operating Instruction (01)
S023-8-1,s " Operation. of: the Backflushable Filter Crud Tank" (Tank T-073).
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(2) PCN to 014 S0h328-11"TransferofSpentResintoPortable Solidification System"'(Tanks T-059 and T-060).
(3) PCN to 01 S023-8-12 " Spent Resin Transfer Operation".
These changes (PCN's) are in accordance with Bechtel Design Change Package, DCP_No. 18-N.
As of this inspection the three tanks requiring vacuum breakers were not included in the licensee's surveillance program. However, the licensee indicated that they will be covered when the design changes are completed.
The inspector is satisfied that the licensee has taken adequate measures to protect against vacuum conditions that could result in tank damage. Therefore this item addressing IEB 80-05 is closed, b.
Bulletin 79-14, Seismic Analysis for As-Built Safety-Related Systems The licensee's response to Bulletin 79-14 was reviewed and is considered adequate.
In addition, the following seismic / design related activities were examined:
(1) QA Program.
Bechtel's service contractor EDS Nuclear, Inc.
provided stress analysis for pressurizer safety and relief lines, main steam lines, and component cooling water system.
The "EDS Nuclear QA Manual" Rev.11, dated August 11, 1977, was reviewed to ensure all applicable 10 CFR 50 Appendix B criteria were included. The following portions of the manual were reviewed in detail and are considered adequate:
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-4-Part I QA Plan, Section 5.0 Design Control 5.1 Drawings 5.2 Design Calculations 5.3 Design Review 5.4 Design Change Control
Part II QA Procedures.
QAP 3.2 Preparation, Review, and Approval of EDS Drawings I 4 QAP 3.3 Design Control.
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QAP 3.7 Internal Design ' Reviews:
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'QAP 5.1 Audits (2)
BechteliAudit of-EDS. $During'the period of the EDS contract, Bechtel perfgrmed three audits of EDS activities including QA program, organization, design control, document control, QA records, and audits. The' audits were performed April 24-25, 1980, January 30-31 and February 1, 1979 and June 4, 1979.
The inspector reviewed these.Bechtel audit reports and found them satisfactory.
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(3) ' Verification of Design C'alculations. The inspector reviewed calculation number MDSC-083 for the relocation of the pressurizer level condensate chamber per design change package 140J.
The calculation sheets documented the seismic analysis and were signed by both designer and checker.
(4) Audits.
The inspector reviewed SCE's audits of design control and as-built verification and found.them satisfactory.
Based on this review and examination of as-builts performed during previous inspections, the inspector considers this item" closed.
c.
.Bulletin 79-15, Deep Draft Pump Deficiencies The licensee's response to Bulletin 79-15 was reviewed and is considered adequate. The response addresses salt water cooling pumps S21413MP112, P113, P114, and P307. The inspector reviewed maintenance and testing data for these pumps and noted no hardware deficiencies similar to those identified in the Bulletin. This item is closed.
4.
Licensee Action on 10 CFR 50.55(e) Reportable Construction Deficiencies The following construction deficiencies were examined by the inspectors for reportability, thoroughness of evaluation, and corrective action:
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'a.
Incorrect Welding Procedure - Reactor Head-Vent The-use of an incorrect welding procedure on weld "A" o'f the reactor vessel vent system was first reported by the licensee under the provisions of 10 CFR.50.55(e).in September 1981. Weld procedure P8-T' Ag, qualified for helding stainless steel to stainless steel (P8 to P8), was used, to weld stainless. steel to inconel (P8 to P43).
The weld has subsequently been cut out.and rewelded in accordance-with'the proper procedure (no. P43-P8-T-Ag). The inspector
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reviewed the weld repair instructions described in Bechtel NCR P-3325 to assure.that the corrective actions were appropriate.
Through discussions with Bechtel welding engineers as to how this condition was discovered the inspector was satisfied that this
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problem was an isolated incident and not a breakdown of the Bechtel quality assurance program. This. item is closed.
b.
Pressurizer Safety Valv'e Operational-Instabilities During EPRI Tests
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The licensee. reported.a condition involving operating instabilities of a-Dresser Industries spring loaded safety valve which were
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exhibited during type, tests ~being conducte'
'lectric Power Research Institute (EPRI). This was repe jer the provisions of 10 CFR 50.55(e) in July;1981. The Dres.. valve tested is
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the same model supplied.for use as pressurizer safety valve on
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San Onofre Units,2 and 3.
Although~the final EPRI report is not due until April 1982 the
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licensee has;used results,from these tests.as guidance for providing the necessary corrective action.to. prevent faulty valve operation.
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EPRI calculations showed that~ a shorter pipe length provided stable.
'F valve operation. Consequently, the licensee has reduced the installed-inlet piping length from,18 feet to 3.6 feet.
Ring adjustments were also made to the valve'to give a higher blowdown pressure
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resulting'in a more stabld valve performance (blowdown is a measure of.the valse's reseat pressure).
Theins'pectorrehiewed'designdrawingS2-1201-ML-032 sheet 1, Rev. 9,'and S2-12010ML-033 sheet 1, Rev. 8.
The latest revisions of the drawings describe the necessary changes for installation of the Unit 2 pr,essurizer safety valves as recommended by the EPRI test progran. The inspector also reviewed Design Change Package (DCP) No.' 24 P/D Rev. O of 9-1-81 which describes the modifications of the' pressurizer relief lines to reduce the length of pipingsbetween pressurizer and safety valves. The required approvals'ere obtained from the plant design supervisor, project w
engineer and quality' assurance. The: inspector examined the Field Welding Checklist for welds H, J,'and K associated with modification k
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e-6-of tte relief line.
Thes'e checklists (travelers) listad the' proper material identification, weld filler metal and weld procedure,
and contained the required QC and ANI approvals.
In conjunction with these checklists the filler metal withdrawal forms were examined
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and found to be compatible with the corresponding field welding checklist. Acceptable liquid penetrant and radiographic examinations were performed following the welding operations; ultrasonic examination performed during preservice examination also showed these welds to be acceptable.
The valve ring adjustment on the pressurizer relief valves for Units 2 and 3 were made by the valve manufacturer, Dresser Industries.
This was done in accordance with Dresser procedure number NC007-81-22 RSH Rev. O of 10-22-81 entitled " Procedure for Adjusting Rings-Model:
31700 Pressurizer Safety Valves".
All work has been completed on Unit 2 including hydrostatic examination.
NCR's have been written and red tags placed on the Unit 3 pressure safety relief valves until the modifications are completed. The inspector is satisfied that the necessary corrective actions have and are being taken to enable stable valve operation. This item is closed.
c.
Core Shroud Guide Lug Insert SLrface Indications
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Cracking of the guide lug inserts (shims) for the reactor internals was reported by the licensee under the provisions of 10 CFR 50.55(e)
in Auoust 1981.
The cracked guide lug inserts are constructed of a base material of Type 348 stainless steel, overlayed with hard facing material.
Combustion Engineering (CE) will remove the existing inserts from service and replace them with Type 304 stainless steel hard faced inserts. The inspector reviewed the
Field Surveillance Report (FSR No. M-410-81 of 10-13-81) verifying proper placement of'the core shroud, guide lug inserts in the Unit 2 reactor vessel internals. Unit 3 replacement is in progress and is scheduled for completion in December 1981. CE has issued drawing number 758-2571 of 9-1-81 entitled " Guide Lug Insert (Replacement)".
The drawing 'shows guide material to be ASTM A 240 Type 304 stainless steel with Stellite 25 hardface. These actions appear adequate to correct this deficiency. This item _is closed.
d.
Excess Bypass Flow to HPSI Pumps The bypass flow orifices have been replaced and the system has been satisfactorily tested. This item is close.-
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Agastat Model 7000 Relays:
Premature Time Out All of the relays with suspected code dates have been replaced.
These include 11 in Unit 2, 15 in Unit 3, and 10 in the warehouse.
The nonconformance reports dealing w)ith the replacement (startup
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NCR 671-E, NCR-E2463, and NCR-E2464 were reviewed and are complete.
This item is closed.
f.
Limitorque Operator Keys - Improper Material The inspector reviewed the Construction Inspection Data Reports for disassembly and reassembly of the valve operators for key replacement for valves HV-9337 and 9339 for both Unit 2 and 3.
The records indicated that the work was complete. This item is closed, g.
Improper Assembly of Coaxial Connectors in Electrical Penetrations The inspector examined two of the reworked connectors in the field to verify that the specified corrective action had been accomplished.
Nonconformance reports E-3124 and 3181 for Unit 2 and E-3215 and 3216 for Unit 3 were reviewed and indicated that all the subject connectors had been reworked and inspected. This item _is closed.
h.
Steam Generator Feedring Deformation The feedring replacement had been completed in both Unit 2 steam generators and was in progress in the Unit 3 steam generators.
The inspector entered the Unit 3 south steam generator and observed one in-process weld, examined two welds where the root pass was complete, and examined one completed weld. The non pressure boundary welds were being performed in accordance with CE welding procedures and quality controls. Weld rod control-for this work was examined ind considered satisfactory. The inspector discussed
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the requirements for weld rod control with two welders who had also welded the Unit 2 steam generator feed rings and was satisfied
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with the controls for these welds. This item is closed.
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Diesel Generator Engine Camshaft Bracket Failure The inspector reviewed the manufacturers correspondence regarding the bracket and the documentation covering the replacement. The
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diesel has been successfully retested. This item is closed.
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Improperly Sized Wire Connectors on Radiation Monitoring Devices The licensee's final reports dated October 23, 1981 specified replacement of wire connectors for the photohelic gas flow gauges with connectors of proper size.
The inspector examined the proper connectors in the following radiation monitoring devices to verify the licensee's corrective action:
Containment Airborne Monitors 2 RT-7804-1 and 2 RT-7807-2 Fuel Handling Vent Airborne Monitors 2 RT-7822-1 and 2 RT-7823-2 Control Room Airborne Monitors 2 RT-7824-1 and 2/3 RT-7825-2 This item is closed.
5.
TMI Modification (Unit 2)
The following TMI modifications, relative to Appendix B of NUREG 0737, were examined:
safety valve indication, containment water level, containment pressure, containment hydrogen concentration, subcooling meter, and pressurizer level' indication from Class 1E Suses. The installations were either substantially completed or completed.
No deviations or items of noncompliance were identified.
6.
Cable Installation Records Review (Unit 2)
Installation records of the following power, control, and instrumentation cabler were reviewed for compliance to applicable procedures and specification:
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a.
Power and Control Cables (1) power cables to charging pumps (2) power cables to.HPSI pumps P017 and P018 (3) power cables to battery chargers 2B001 and 2B003 (4) power cable to DG exhaust fan (5) control cables to SI valves 6) control cables to HPSI pumps P017 and P018 7) control cables to LPSI pumps P015 and P016 8) control cables to containment spray pumps P012 and P013 b.
Instrumentation Cables (1) High linear power level
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(2) High containment pressure
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(3) Low pressurizer pressure
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(4) Containment isolation (5) Containment combustible gas control (H recombiner)
(6) Safety injection
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-9-(7) Auxiliary Feedwater (8) Shutdown cooling system (9) Loop 1 hot leg temperature (10)Pressurizersafetyvalvepositionindication (11)S/G1 level (12) Auxiliary feedwater flow (13) Refueling water tank (T005) level (14) Radiation monitoring system (15) Hp gas concentration (16)Subcoolingmeter (17) Containment area sump level No deviations or items of noncompliance were identified.
7.
General Unit 2 Tour, During inspection tour of Unit 2 the inspector noted that, for LPSI pump number P016, Class IE conduit AVAR26 for the sump. level alann switch was unsupported for a length of approximately 14 feet. Review of the conduit inspection record indicated that the cotiduit installation (including supports) had been inspected and accepted on May 5, 1980.
Further investigation indicated that subsequent to the conduit inspection, the conduit. support had been removed to facilitate installation of structural _ steel ; associated with pipe support S2-SI-009-H004, Rev. 3, DCN 8 dated June 16, 1981.
Bechtel work plan procedure WPP-037, "Interdiscipline Notification for Plant Item Removal Requirements," paragraph 4.1 requires that a Plant Item Removal Request Form (PIRR) be initiated when an item or component in the jurisdiction,of another discipline requires disassembly or removal. "No PIRR had been' written for removal of the conduit support.
Failureloproperlydocumenttheremovalofcompletedworkisconsidered an item of noncompliance (50-361/81-34/01).
8.
Exit Meetings The inspectors met with licensee representatives (denoted in paragraph 1)
on November 20 and December 11, 1981. The scope of the inspection and findings as detailed in this report were discussed.
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