IR 05000361/1990015
| ML20043H700 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/11/1990 |
| From: | Huey F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20043H697 | List: |
| References | |
| 50-361-90-15, 50-362-90-15, NUDOCS 9006260314 | |
| Download: ML20043H700 (8) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-361/90-15 and 50-362/90-15
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License Nos.
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Licensee: Southern California Edison Company
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23 Parker Street Irvine, California 92718 Facility Name: San Onofre Nuclear Generating Station
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Inspection at: San Clemente, California Inspection Conducted:
April 23 - May 11, 1990 l
Inspectors:
F. Gee, Reactor Inspector
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W. Wagner, Reactor Inspector
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K. Joanston, Resident Inspector 0. Corporandy, Reactor Inspector
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Approved by F
[o!//!Qh F. R. Huey, Chief Date ligned
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Engineering Section
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Summary:
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Inspection During the Period of April 23 - May 11, 1990 (Report
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Nos. 50-361/90-15 and 50-362/90-15)
l Areas Inspected:
A special unannounced inspection by regional inspectors of
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the licensee's design and engineering programs.
Inspection procedures 30703, 37700, 37701, 62700, 73753, 92700, 92701 and 92702 were used as guidance for the inspection.
Results:
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General Conclusions and Specific Findings
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During the review of design change packages, several engineering weaknesses were identified:
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Inadequate Engineering Review a.
An error associated with fluctuation of containment pressure was not
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accounted for in the setpoint calculation for Anticipated Transients
Without Scram (ATWS)/ Diverse Scram System (DSS).
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Multiple examples of inadequate translation of design change requirements from Unit 2 to Unit 3 were noted by the inspectors.
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9006260314 900611 PDR ADOCK 05000361 Q
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Inadequate Engineering Design Implementation j
Signals from the ATWS/ DSS were not compatible with the existing Critical Function Monitoring System.
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Inadequate Engineering Design Control Containment location drawings and terminal point assignments were not adequately controlled.
Significant Safety Matters:
None Summary of Violations or Deviations:
None Open Items Summary:
Three open items were closed, and one open item remained
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open.
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DETAILS 1.
Persons Contacted Southern California Edison
- R. Baker, Project Engineer
- R. Borden, QA En ineer
+C. Brandt, QA En ineer
+D. Brevig, Onsit Nuclear Licensing (ONL) Supervisor
+*R. Bridenbecker, Vice President, Site Manager M. Cabrera, Construction Engineer
+*A. Brough, Engineering Supervisor
+L. Cash, Maintenance Manager J. Chang, Station Technical Engineer
+R. Clar(, Station Technical Engineering Supervisor F. Chiu, Controls Engineer T. Elkins, Construction Supervisor K. Flynn Station Technical Mechanical Engineer S. Gensha,w, Maintenance R. Haverkamp, Construction Engineer
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J. Kan, Controls Engineer
- S. McMahan, Maintenance
+H. Morgan, Station Manager
+J. Patterson, Assistant Manager Maintenance Units 2/3
+D. Peacor, Manager, Station ;mergency Preparedness
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Pla Station Technical ert, Station Technical Cognizant Engineer
+N.
Qui ey,
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+P. Sha er, Compliance Supervisor
- M. Short, Station Technical Manager
+*K Slagle, Deputy Station Manager
- R. St. On
+R Waldo,ge, Controls Supervisor Assistant Technical Manager P. Watson, Compliance Engineering Su)ervisor
+*D. Werntz, Onsite Nuclear Licensing Engineer San Diego Gas and Electric
+*R. Erickson, Senior Engineer NRC
+*A. Hon, Resident Inspector The inspectors also held discussions with other licensee and contractor personnel during the course of the inspection.
4 Attended the Exit Meeting on May 4, 1990
- Attended the Exit Meeting on May 11, 1990 m
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2.
Review of Design Change Packages (DCP's) (37701)
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Several engineering weaknesses were identified during review of recently l
implemented design change packages.
l a.
Inadequate Engineering Review (1)CalculationJC-EGA-003,apartofDCP 2/3-6553 Anticipated
Transients Without Scram (ATWS)/ Diverse Scram System (DSS){he was e
used to determine the system loop error and to verify that
diverse scram trip setpoint of 2450 psia met the setpoint
requirements of Combustion Engineering's " Functional Design t
S)ecification for the Diverse Scram System for Compliance with tie ATWS Rule 10 CFR 50.62".
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The calculation did not account for the error contribu'ted by fluctuation of containment pressure.
The assumption made in the calculation, that containment pressure was constant, was
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not verified.
Four qualified gauge pressure transmitters were
used in place of four absolute pressure transmitters in monitoring pressurizer pressure.
The scale used in the measurement was in units of absolute pressure.
An offset of
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14.7 psi was calibrated into the gauge transmitter.- Fluctuations
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in containment pressure, which ranged from.+1.5 psig to -0.3 psig according to Unit 2'and 3 Technical Specifications Section 3.6.1.4, would bias the trip set)oint. - The resultant upper and lower setpoint limits would )e 41.5 psi higher and -0.3 psi lower than what were shown in the calculation for upper and lower setpoint limits respectively.
The inspector confirmed that
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the ATWS/ DSS setpoint in the calculation included sufficient margin (10.8 psi) to absorb the unaccounted error without causing
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an overlap with the lower margin of the lifting setpoint of the
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i pressurizer safety valves.
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The calculation, issued on October 18, 1989,by two levels ofhad been revi an independent review engineer and approved supervisory engineers.
Yet the inaccuracy contributed to the scram trip setpoint by the fluctuation of containment pressure was not discovered until noted by the NRC inspector.-
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(2) Inadequate engineering reviews were noted associated with conversion of design change packages from Unit 2 to Unit 3.
Specifically:
In.the fault current calculation on page 1651 of DCP 3-6674.00BJ (Moratta Valve Replacement, Revision 0), a Unit-2 fault current numerical value had not been deleted from the Unit 3 package.
A different cable length contributed to a higher a
fault current value for Unit 3, and the higher fault current was
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not properly identified in the Unit 3 package.
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A similar finding was reported in Inspection Report 90-14, in that test guidelines for check valves were carried over from
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Unit 2 to Unit 3 (MMP 3-6753.0SM), when these check valves were being removed from Unit 3.
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b.
Inadequate Engineering Design Implementation Inadequate engineering d,tsign implementation was observed in
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DCP 6653, ATWS/DS$.
Specifically, signal compatibility of the ATWS modification with the existing Critical Function Monitoring System t
(CFMS) was not well thought through by design engineering.
The ATWS system included a 14-bit analog-to-digital conversion card while the CFMS utilized a.12-bit conversion.
Also, the ATWS modification i
used a voltage signal of 15 volts while the CFMS used a 8.3 volts signal.
A review of signal compatibility should have been performed by design engineers prior to equipment purchase rather;
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than after the equipment had been delivered to the site.
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c.
Inadequate Engineering Design Control.
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(1) Inadequate engineerina design control was noted in DCP 6653 ATWS/ DSS.
During implementation of the DCP, three of four
pressure transmitters had to be relocated to different locations
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within containment since the designated locations had already
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been taken by other equipment.
Containment location drawings i
had not been maintained to reflect as-built conditions.
(2) Inadequate control of terminal point assignments were also r
noted during review of various design change packages.
Terminal points, thought to be spare and assigned for use in a DCP, had actually been used for other design change packages.' As noted
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above, proper drawing configuration control had not been
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maintained.
No violations or deviations were identified.
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3.
InserviceInspection-ObservationofWorkandWorkActivities(73753,}
The Unit 3 inservice inspection program was based on the requirements of
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the ASME Boiler and Pressure Vessel Code,Section XI, 1977 Edition.
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witn addenda through the summer of 1979.
During this inspection, the
licensee was conducting the Unit 3, Cycle 5, refueling outage' which was
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the first refueling outage of the second period of the first_ ten year inservice inspection interval,
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The inspector observed ultrasonic examinations being performed on the following four areas of the low pressure safety injection-(LPSI) header,
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S3-1204-8"-C-FE0:
ISI ID Number Examination Area Description f
03-073-1850 Penetration No. 49-to-8" Sch 140 pipe
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03-073-1860 8" Sch 140 pi)e-to-elbow-03-073-1870 8" Sch 140 el)ow-to-pipe 03-073-1880 8" Sch 140 pipe-to-valve
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l The following attributes were evaluated and found to be consistent with
the approved procedure and ASME Section XI requirements: the type of
a pparatus used, scanning technique, extent of coverage, calibration of
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tie instrumentation and system prior to examination, beam angles, size
and frequency of the search unit limits of evaluation and recording of indications, and determination of acceptance limits, In Conjunction with this examination, the inspector reviewed the calibration block r
certifications, transducer RF pulse wave forms, and the ultrasonic
couplant certifications, j
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t The inspector also reviewed the qualification and certification records
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of the Level II and III NDE personnel performing the inservice
inspection examinations. The records confirmed that the examiners were
qualified within the guidelines of SNT-TC-1A.
No violations or deviations were identified.
4.
Procedural Deficiency During Main Steam Safety Valve Ring Setting t
(37700, 62700)
I The inspector observed portions of the implementation of design change l
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package MMP 3-6806.0SM, Main Steam Safety Valve Ring Settings.
DuringobservationofthesettingoftheMainSteamSafetyValve(MSSV)
nozzle and guide rings, the inspector noted that maintenance personnel
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werenotimplementingallaspectsoftheMaintenanceOrder(MO)and
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referenced procedure.
In particular:
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In step 6.2, the M0 for each of the 13 MSSVs stated that it might be necessary to remove the discharge tailpiece of the MSSV to perform the
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ring adjustments.
Furthermore, it stated that if this was the case, it
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would be necessary to remove and replace the discharge pipe in accordance l
with steps 6.1.1 and 6.5.6.8 of maintenance prccedure S023-I-6.45 i
Mains' team Safety Valve Inspection, Repair and Calibration." Step 6.5.6
described the steps for reassembly of the discharge flange, including the
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use of a new flexitallic gasket. The step also required-that the initial i
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l gasket thickness and final gasket crush be recorded on a maintenance data
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recordform(MDRF). The inspector noted that this information was not
being recorded.
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When questioned as to why an MDRF was not being used to record gasket information, maintenance personnel responded that M0 step 1, stated that i
procedure S023-I-6.45 was to be used as a reference guide only and that this was interpreted to mean that an MDRF need not be filled out. The
inspector reviewed this item with the Unit 2/3 maintenance manager.
l The manager concurred with the inspector that an MDRF should have been
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used. The manager noted that an MDRF had not been included in the M0
't because step 6.2 had been identified as optional.
He stated that corrective action would be initiated to address the clarity of M0
l instructions regarding the use of referenced procedures.
In addition, he
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comitted to evaluate the need to enhance training for maintenance
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personnel.
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5.
(Closed) Followup of Licensee Event Report 50-361/89-10 (92700)
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The inspector reviewed LER 50-361/89-10, including Revision I concerning the MSSV nozzle and guide ring settings.
Basedonareviewofthe
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associated non-conformance report, desi'n change, and observation of the implemented corrective actions, this LE is closed.
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6.
Licensee Action on Previously Identified Items (92701, 92702)
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(Closed) Followup Item No. 50-361/88-35-01 Failure of the Licensee i
to Incorporate Design Criteria for Emergency Chi:ler Freon Levels i
into Appropriate Station Operating and Paintenance Instructions (92702)
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(1)CalculationEC-119, Revision 0,fordeterminingappbriate freon levels was reviewed by the inspector.
Tae ca lation appeared to consider appropriate design parameters in t
establishing acceptance criteria for rean fluid levels.
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i (2) Surveillance Procedure 50H3-0-9, Revisits O, " Operator Rounds
and Inspections" was reviewed by the inspector.
Review of the
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l inspection procedure pertaining to emergency. chiller Freon l
1evels appeared to demonstrate compliance with the acceptable
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Freon levels established in EC-119.
(3)SONGSMaintananceordarNo. 90050773000 was reviewed by the
inspector.
This maintenance order provided guidance on checking acceptable Freon levels during emergency chiller operation and included instructions for adding Freon to maintain required
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levels.
Per this maintenance order (MO) Freon level during o)erationwouldbeobservedtobeapproxImately1inchlower.
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chillers were not operating.
Inconsistent with this, the
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allowable observed maximum Freon level given in the M0 was not i
adjustedtobe1inchlowerthanforthecorrespondinglevelto be observed during non-operation.
The inconsistency of the subject MO was later corrected by the licensee, b.
(Closed) Followup Item No. 50-361/88-35-02 Failure of the Licensee
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to Take Prompt Corrective Action on the Deficiencies due to Low
Emergency Chiller Freon Levels (92702)
s (1)Asdiscussedabove,itappearedthatdesigninformationhadbeen i
adequately translated into operating procedures and instructions such that deviations from required Freon levels could be promptly identified.
Review of the o>erating procedures and instructions identified above confirmed t1at-instructions on appropriate
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corrective actions also appeared to be adequate.
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(2) The completed corrective actions regarding prompt initiation of NCR's as discussed in the July 21, 1989 SCE letter from Kenneth P. Baskin to the NRC.were reviewed by the inspector and appeared satisfactory.
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(0 pen) Followup Item No. 50-361/88-35-03 Emergency Chiller System i
Heat Loading Calculations Which Contain Errors and lack of i
Requirements to Periodically Monitor Performance of Emergency I
Chilled Water System Components (92701)
(1)CalculationM-73-61, Revision 4,"EngineeringSafetyFeature(ESF)
Switchgear (SWGR) Room Heat Load Calculation was reviewed by the inspector.
Inconsistencies identified during inspections 50-361/88-35 and 50-362/88-37 were still not corrected.
Review ofthesubjectcalculationdemonstratedthattheneteffect of the inconsistencies appeared to render Calculation M-73-61 more conservative.
(2)CalculationM-75-50, Revision 0," Safety'EquipmentBuildingHeat Loads - Emergency" was reviewed by the inspector.
The inspector performed a walkdown of room 005 in Unit 2 and reviewed the heat load calculation for room 005.
The calculation appeared to reflect the appropriate design basis.
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(3)Attheexitmeetingtheinspectornotedthattheerrorinthe emergency chiller heat load calculation was identified last year At that time, SCE had acknowledged that other emergency chiller
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heat load calculations may also include design basis errors.
The inspector expressed concern that errors in tie heat load calculations could have provided the potential for underestimated
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heat loads.
The inspector noted that the SONGS Commitment Register (SOCR) committed to a December 15 1990 date for reconcilingemergencychillerheatloadcalculationswiththe design basis.
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(4)Theinspectordidnotobserveanyprogressontheissueof performance monitoring of the emergency chilled water system components.
The inspector noted a July 12, 1990 SOCR commitment to develop a plan for a performance monitoring system for the
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emergency chiller.
The licensee acknowledged that the plan would be developed by July 12, 1990.
This item remains open.
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7.
Exit Meeting (30703)
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The inspectors conducted exit meetings on May 4 and 11, 1990, with members of the licensee staff as indicated in paragraph 1.
During these meetings, thr, inspectors summarized the scope of tie inspection activities aid reviewed the inspection findings as described in this report.
The licensee acknowledged the concerns identified in the report,
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