IR 05000361/1990014

Insp Repts 50-361/90-14 & 50-362/90-14 on 900402-20 & 0430- 0504.No Violations Noted.Major Areas Inspected:Licensee Design,Engineering & Associated Quality Verification Activities

ML20043D723
Person / Time
Site:	San Onofre
Issue date:	05/24/1990
From:	Huey F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20043D720	List: IR 05000361/1990014 ML20043D719
References
50-361-90-14, 50-362-90-14, IEB-89-002, IEB-89-2, NUDOCS 9006110110
Download: ML20043D723 (13)

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U. S. NUCLEAR REGULATORY COMMISSION

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REGION V

Report Nos.

50-361/90-14 and 50-362/90-14 License Nos.

NPF-10 and NPF-15 Licensee: Southern California Edison Company 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 2-20 and April 30 - May 4, 1990 Inspectors:

D. Corporandy, Reactor Inspector F. Gee, Reactor Inspector K. Johnston, Resident Inspector M. Miller, Reactor Inspector W. Wanner, Mct r Ir 59ector

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5~#'/!fo Approved by:

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s F. R. Huey, Chief

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Date Signed Engineering Section

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Summary:

i Inspection During the Period of April 2-20, and April 30 - May 4, 1990 (Report Nos. 50-361/90-14 and 50-362/90-14)

Areas Inspected:

A special unannounced inspection by regional based inspectors of the licensee's design, engineering and associated quality verification activities.

Inspection procedures 30703, 35702, 37700, 37701 and 37702 were used as guidance for the inspection, Results:

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General Conclusions and Specific Findings The licensee's Nuclear'0versight Division has shifted emphasis from the traditional QA programmatic compliance audits to performance based assessments that observe activities which have an impact on plant safety and reliability.

A number of concerns were identified in the following areas:

the need to include Atmospheric Dump Valves (ADV) nitrogen backup Pressure Control Valves (PCV) in the Inservice Testing-(IST) program; the effects of degraded o

instrument air pressure on the oaeration of the ADVs was not assessed; lack of L

post modification walkdowns by t1e environmental qualification group; lack of full understanding of site operations and policies by corporate engineering personnel; inadequate response to IE8 89-02; and untimely issue of a Part 21 report on the deterioration of the snubber grease.

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Significant Safety Matters:

None n,

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Summary of Violations or Deviations:

None.

Open Items Summary:

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One open item was identified, L

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DETAILS 1.

Persons Contacted Southern California Edison

+*D. Brevig, Nuclear Engineering and Design ManagerOnsite Nuclear Lice

• M. Merlo,

• J. Reilly, Station Technical Manager

• B. Katz, Nuclear Oversight Division Manager

+*H. Morgan, Station Manager

• D. Herbst, Site QA Manager

+*D. Werntz, ONL Engineer

• M.Short,ProjectManager,DesignBasisDocuments

• M..Speer, ONL Engineer

• R. Plappert, Compliance Supervisor

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+*P. Shaffer, Compliance Supervisor

+*C Brandt, OA Engineer

• I Katter, Hechanical Supervisor

• G. Gibson, ONL Engineer

+*R. Baker, Project Engineer

• T. Elkins, Construction-Supervisor

• J. Wambold, Project Manager

• R. Reinsch, Bechtel QA Manager

• R. Strom, ISEG Supervisor

• A. Thiel, Electrical Supervisor

• A. Kaneko, Electrical Supervisor

+*A. Brough, Site Engineering Supervisor

• N. Maringas, QA Engineer

+*R. Bridenbecker, Vice President, Site Manager

+ N. Quigley, Station Technical Cognizant Engineer

+ D. Peacor, Manager, Station Emergency Preparedness

+ L. Cash M

+K.Slagie,aintenanceManager L

Deputy Station Manager

+ R. Clark, Station Technical Engineering Supervisor

+ J. Patterson, Assistant Manager Maintenance Units 2/3

+ R. Waldo, Assistant Technical Manager M. Herschthal, Station Technical Lead Engineer

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H. Chun, Quality Engineering Supervisor C. Chiu, Safety Engineering Manager C. Olvera, Quality Engineer G. Mcdonald, Quality / Surveillance Supervisor Programs Supervisor l

R. McWey, Evaluation L

M. Cave, Procurement Agent l

J. Harmon, QA Engineer F. Chiu Cont W.Wilkinson,rolsEnineer Mechan cal Engineer R. Wise, EQ Electrical Engineer W. Nakamoto, EQ Electrical Engineer V. Thomas, EQ Electrical Engineer

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L S; Root, Nuclear Supervisor, Integrated Plant Review C. Diamond, Mechanical Engineer

.N. Basilio, Electrical Engineer San Diego Gas and Electric

+*R. Erickson, Senior Engineer

• R. Lacy, Vice President

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NRC

• C. Caldwell, Senior Resident Inspector

• R. Huey, Chief, Engineering Section

+ A. Hon, Resident Inspector The inspeuors also held discussions with other licensee and contractor personnel during the course of the inspection.

• Attended the Exit Meeting on April 20,1990

+ Attended the Exit Meeting on May 4, 1990 2.

Quality Verification Function (35702)

The Nuclear Oversight Division has the responsibility for evaluating compliance with the Design Prog'am which is accomplished, in part, by audit and surveillance of quality-affecting activities.

Of specific

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interest during this inspection was the extent of licensee QA oversight involvement with in-house design and engineering activities, significance of the findings, and conclusions regarding the quality of the overall engineering design effort.

The following is a list of audits and surveillances performed during the last two years related specifically to design activities.

(a) Surveillance Reports GOS No.

Report Date Subject

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2-88 1/21/88 Design Interface Control

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7-88 2/2/88 Design Calculation Changes 12-88 3/31/88 Design Interface Control n.-

14-88 6/30/88 Configuration Control 15-88 5/9/88'

Design Change Package Control 24-88 7/1/88 Backup Nitrogen System Design Bases and-Criteria 26-88 8/8/88 Drawing Control 31-88 11/15/88 Design Interfaces

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37-88 12/5/88 Design Calculations f

2-89 2/17/89 Field Change Control

11-89 4/26/89 Design Calculations 15-89 6/8/89 DCPs/LCPs

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18-89 6/15/89 Supplier Document Control

22-89 Report-in Design Calculations progress

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31-89 Report in Outage Surveillance progress

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Audit Reports Report No.

Report Date Subject

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BPC-2-89 12/18/89 Bechtel Design Activities SCES-029-89 12/27/89 NEDO Design Change Process Review of these reports indicated that the audits and surveillances were performed by, qualified personnel, with the findings clearly documented.

Of greater significance, the reports revealed a change of emphasis from the traditional programmatic compliance audits to performance-based audits, which provide assessments of areas important to plant reliability

.and' safety.

The licensee has initiated a performance-based assessment technique that-is separate from their approved QA program but which is designed to I

enhance the Nuclear Oversight Organization s ability to identify performance degrading conditions before they become significant problems.

This technique is entitled " Vertical Assessments of Design and Design Change Packages".

The first vertical assessment was to verify the

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adequacy of selected aspects of the design changes included in DCP 3-6674.00BJ Main Steam Isolation Valve (MSIV), Main Feedwater'(MFW)

Isolation Valve and MFW Block Valve Modifications.

Deficiencies were identified in design calculations and drawings, processing of design calculations, and of the six assessm, processing of EQ package changes.The qualifications ent' team members were reviewed; all had the necessary experience required to perform the assessment.

In addition, all team members received special training on the concepts of performance-based assessment techniques and how to plan, conduct and document the assessment.-

No violations or deviations were identified.

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Review of Design Change Packages (DCP's) (37701)

Walkdowns were performed on the following Unit 2 and 3 design change packages:

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DCP 6005.02 ADV Modifications

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DCP 6605.1 Control Room Human Factors Modifications c.

DCP 6605.4 Remove Reactor Regulation System

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DCP 6605.5 Control Room Human Factors Common Panels e.

DCP 6553 Anticipated Transients Without Scram / Diverse Scram System f.

DCP 6674 Marotta Valve Replacement

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DCP 6788 Removal of Control Element Drive Mechanism (CEDM) Snubbers h.

ItMP 6392.00SE Addition of Local Annunciator for Non-1E Inverter i.

MP 6753 ADV Backup Nitrogen Check Valve Test Connections The inspectors noted the following comments and concerns:

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DCP 2/3-6674.00BJ - Marotta Valve Replacement The purpose of this DCP was to upgrade the control systems of the main steam isolation valves, the main feedwater isolation valves, the main feedwater block valves and their associated hydraulic power units to increase reliability so that a single failure would not result in a spurious valve closure.

Revision 0 of the original DCP was supplemented by a large number of Field Interim Design Change Notices and Retrofit Problem leports.

The large number of change notices and problem reports appeared to indicate:

Incomplete design work.

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Inadequate coordination between engineering and operations

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as evidenced by the implementation of Sollex bypass and curren,t monitor i

by field interim design change notices. (Sollex was a device to-prolong the life span of the solenoid coil by reducing the sustained voltage to the coil.)

Inadequate engineering review in which a signal path was allowed to

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reach the pump start circuit from the alarm circuit, causing an inadvertent pump start.

On April-17, 1990, during a walkdown of the Unit 3 MSIV's associated with

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review of DCP-6674.00BJ, the following observations were made by the

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inspector regarding environmental qualification (EQ) concerns:

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screws were loose on some electrical pull boxes

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a flexible metal hose covering an electrical.line to the NAMC0 limit

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switches appeared to have an unacceptably sharp bend radius some terminal boxes were waterproofed on one side but not on the

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other Further investigation revealed that the flexible metal hose was not required for the electrical line to the NAMC0 limit switches and that the j

subject terminal ocxes did not require waterproofing.

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This inconsistent application of EQ requirements and the lack af any post-modification walkdown by environmental qualification engineers is I

indicative of inadequate engineering involvement in plant activ! ties associated with the installation of this design change.

A May 9, 1986 reliability analysis on the configuration of the MSIV dump valve confirmed that the proposed four valve configuration was more i

reliable with respect to the MSIV remaining open during normal operation.

However, the analysis also concluded that the risk of failure of the isolatbn valve to close on demand had increased by a factor of four.

Although the overall risk of failure to close on demand was small, the inspector coted that the safety analysis performed for the DCP did not address the nigher probability that the MSIV and FWIV may fail to close on demand as a result of solenoid valve failure.

In addition, DCP 6621, which earlier replaced the two dump valve configuration for the MSIV's with a similar four vcive configuration, did not address the higher probabilistic risk of failure of the MSIV's to close on demand as a result of failure of a solenoid valve.

The inspector noted that the safety analysis for an earlier version of DCP 6621, entitled Proposed Facility Change (PFC) 6621, discussed the higher probability of failure of the MSIV to close on demand due to failure of

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the solenoid valves.

The information in this safety analysis apparently

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was not carried over frcm the PFC to the design change package.

In discussions with the Nuclear Safety Group (NSG), it was clear that the risk of failure to close on demand had also been analyzed in an NSG probabilistic risk assessment dated April 6, 1990.

It appeared that DCPs-6621 and 6674 safety evaluations did not

with the Nuclear Safety Analysis Center (provide information consistent NSAC) guidelines.

The implementation of NSAC 125 guidelines is discussed later in this report.

Also, the DCP safety evaluations did not provide information concerning accident probability which was already available in licensee documents.

DCP 6713, Fuel Handling Isolation System (FHIS) Condensation Collectors

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This DCP added condensate collectors to the FHIS radiation monitor i

sampling line.

The licensee determined that the humidity generated by the spent fuel pool during a core offload c W ensed in the sampling lines and caused spurious actuations of the FHIS. "The inspector reviewed the DCP to evaluate engineering adequacy and made the following observations.

The inspector noted that the radiological safety requirements in the design criteria stated that health physics personnel would install berms around the condensate pots.

During a walkdown, the inspector noted that, although the modification had been completed and turned over to the plant, berms had not been installed.

Although the lack of a berm is of low safety concern, the turnover procedure required that the DCP be verified to have been installed as documented.

The inspector considers the DCP requirements for a berm should have been implemented as described in the DCP, or an appropriate change should have been obtained.

Licensee engineering management determined that the failure to install the berm had two causes.

First, corporate engineering personnel were unfamiliar with the site policy for berms, and the DCf was approved imposing a

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requirement which was not in accordance with site policy.

Second, after construction, the turnover of the installation to the site took the health abysics berm policy into account and accepted the DCP as complete, althoug1 the required berm was not installed.

Licensee engineering and construction management stated that they would use this issue as a lesson learned to emphasize the need for corporate engineers to understand site operations and policies, and to emphasize the need for turnover engineers to verify installation of DCPs in a more rigorous manner.

DCP 6788, Removal of Control Element Drive Mechanism (CEDM) Snubbers.

This DCP removed CEDM snubbers and re part of a snubber reduction program. placed them with rigid struts as The inspector reviewed the design change package (DCP) to evaluate engineering adequacy and made the following observations.

During a technical specification snubber test for Unit 2, several snubbers failed.

The licensee determined that the failures were due to deterioration of the grease which lubricates the snubbers.

The grease was found discolored and solidified.

As a result of these failures, an operability ana<ysis was performed for Unit 3 operating at ThelicenseedeterminedthatthedeterIoratedgrease,poweratthat time.

made by Bel was intended for only fcar or five years of service. Ray, was not intended but The licensee pursuing further documentation of the procurement of the gicase.

To date the licensee has been unable to obtain a record of the quallficationofthegreaseaerformedbyeitherPacificScientific,the snubbermanufacturer,orComaustionEngineeringIedfortiesnubbersdo the sup)1ier of the snubbers,o the licensee.

Vendor manuals su3p1 not address a limit of the service life of tie grease.

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The solidification of the grease resulted in the snubbers being unable to perform their intended function.

However, 3 CE report demonstrated that the locked snubbers would perform adegattly to maintain system safety.

The original design calculation called for the snubbers to allow for

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thermal expansion.

The functional tests clearly demonstrated they did not do this.

The iicensee stated that the su) ply of defective grease had been sufficiently investigated to determine t1at the only licensees who

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had been supplied with this grease were SCE and WNP-3, with whom the i

licensee had communicated details of the failue.

The inspector considered that, without confirmatory documentation from the involved suppliers, this determination was not valid and a 10 CFR 21 report would have been warranted.

Concerning operability of the snubbers, the inspector observed internal licensee correspondence stating that the snubber were at no time non-operational.

This statement was inaccurate since many of these snubbers with this grease failed the technical saecification functional testing.

After discussions with the licensee, tie inspector determined that the licensee may have confused the operability of the component (snubber)withtheoperabilityofthesystem.

Since the document was a justification for not issuing a 10 CFR 21 report, the inspector discussed

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the operability and reportability associated with '.hese failures with the

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licensee during the inspection and the exit meeting

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During the investigation of the failed snubbers, the licensee determined that the snubbers had been sprayed during a head venting operation.

As a result, the )rocedure for venting the head was changed to require the snubbers to )e cleaned and dried if sprayed.

The inspector reviewed the snubber vendor's procedure and the instructions for care of the snubbers after spraying.

The SCE procedure ap

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requirements specified by the vendor.peared to include all the

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No violations or deviations were identified.

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Desion, Design Changes and Modifications (37700) DCP 3-6605.02 - ADV

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Background

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Units 2 and 3 each have two atmospheric dump valves (ADVs), one for each main steam line, which can be used to control steam )ressure in the event the condenser steam dumps are not available.

T1e ADVs are air operated with normal air supplied by the instrument air system and backup safety related air supplied by nitrogen accumulators.

The ADVs require 80 psi to fully open and will fail closed on loss of air pressure.

The following description uses the equipment numbers for the air supplying ADV 8419.

Prior to the implementation of modification MMP 3-6753.0SM(ADV Backup Nitrogen Check Valve Test Connections), the nitrogen to the ADV was supplied from a high pressure bottle, through a pressure regulatorsetat80 psi,throughanairoperatedshutoffvalve(PCV 8454), through a check valve (MUO26) and to an air supply header.

Instrument air, which is normall supplied through a check valve (y at approximately 110 psi, was MUO27) to the header.

The header then supplied air, through an 80 psi air regulator, to the ADV operator.

The bonnet of the actuator for PuV 8454 was supplied with instrument air upstream of check valve MU027.

The purpose of PCV 8454 was to isolate the nitrogen supply during normal operation and to open upon the inss of instrument air.

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The modification MMP 3-6753.0SM was initiated to accomplish the following:

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Install isolation valves and vent connections around MUO27, This would allow the check valve to be leak tested.

This modification was initiated in response to a concern discussed in NRC inspection report 50-361/88-03 regarding the need to test check valves which prevent the loss of safety related backup air in the event of a loss of instrument air.

Replace check valve MUO27 with a more reliable model.

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Delete check valve MU026.

It was determined that this valve was difficult to leak check and was redundant in function to

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PCV 8454.

This portion of the modification was not performed on Unit 2.

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Add a vent valve to the instrument line supplying PCV 8454.

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Design Application of 3PCV-8454 and 3PCV-846.0 1he inspector questioned whether the licensee had addressed the consequences of a degradation but not a complete loss, of instrumentairpressureto3POV-8454.

In this scenario, instrument air would not be available to provide full pressure to open the -

ADVs.

However, instrument air may still maintain 3PCV 8454 partially closed, possibly limiting the ability of the nitrogen

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system to operate tne ADV.

This concern was brought to the attention of the cognizant engineer who concurred that the application of 3PCV 8454 mig 1t degrade the ADV nitrogen backup system during a reduced instrument air system pressure transient.

The cognizant Engineer initiated action to

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determine the design basis for 3PCVs 8454 and 8460 (the corresponding valve for ADV 8421) and the effects of degraded instrument air pressure on the operation of the ADVs. Additionally, action was initiated to provide operators with instructions which address ADV operation under degraded instrument air conditions.

This issue will be carried as an open item pending the results of the licensee's review of the air system design and the effects of the as-found configuration on the operation of the ADVs (50-362/90-14-01).

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These pressure control valves, which are required by the system design to chan minimum value,ge position when instrument air pressure drops to a are not included in the licensee's ASME Section XI Inservice Testing (IST) program.

The licensee agreed to evaluate the necessity of including the PCVs into their IST program.

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Post Modification Testing of the ADV Nitrogen System Modifications e

Section 7 of MMP 3-6753.0SM included testing procedure guidelines for the modifications to the ADV operator air supply system.

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test guidelines requi' red the testing of two check valves which had been removed during the design change and re'placed with a straight length of piae.

The check valves (MUO26 and MUO32) were removed from Unit 3 )ecause they could not be adequately leak tested and it was determined that 3PCV 8454 and 3PCV 8460 performed the same design function.

While this portion of MMP 3-6753.0SM had not been aerformed on Unit 2, the testing portion of the Unit 2 modific6 ion lad been duplicated for the Unit 3 modification.

Theobjectiveco test the check valves was not removed from the test guidelines.

At the time the discrepancy was noted by the inspector, the CWO to implement the test guidelines was still in engineering review and had not been issued for use.

Additionally, had the test oeen performed as written, the discrepancy would have been self

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revealing.

However, this finding indicated a lack of attention to detail when the Unit 2 design change was modified for use on Unit 3.

These discrepancies were identified to the Cognizant Engineer for corrective action.

No violations or deviations were identified.

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Licensee Response to Inspection and Enforcement Bulletin 89-02, Inspection of 5afety Relateo cneck valves (3//00).

The Bulletin required licensees to review all safety related check valves and inspect those with internal fasteners made of 410 stainless steel.

This inspection was requested as a result of failures of this type of fastener in check valves.

The inspector reviewed the licensee's methodology for review and evaluation of the safety related check valves and made the following observations.

For the review of Units 2 and 3 safety related check valves, the licensee reviewed the check valves in the Inservice Testing (IST) program.

The licensee was not aware that the IST program did not contain all the safety related check valves.

The licensee response to the NRC stated inaccurately that all safety related check valves had been reviewed as requested in the. Bulletin.

The inspector noted that the group which prepared the res)onse is the Independent Safety Evaluation Group (ISEG).

To :)repare tie response, ISEG performed the engineering review and then oatained concurrence from other engineering groups.

The adequacy of this methodology is being evaluated by the resident inspectors.

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For the review of Unit I safety-related check valves, the licensee determined that additional safety-related check valves existed which were not listed in the IST program.

These valves are safety related only because the pressure boundary must maintain integrity; the internals of these valves have no safety related function.

Tae licensee refers to these valves as Class 3 safety related check valves.

The licensee decided that Class 3 valves did not require inspection since breakage of a fastener which could impair the internal function would not be of consequence to a safety function.

The NRC responded by stating that all safety related check valves should be inspected.

The licensee methodology for excluding the Class 3 check valves did not address the possibility that the failure of the fasteners could result in fastener parts becoming entrained in the fluid stream and affecting the safety related function of other equipment.

This appears to be an oversight in the engineering area.

In addition, the licensee noted that they have no controlled vendor drawings for valves manufactured by the Crane Company.

Those valves are reviewed by inspection of the vendor catalogs to determine the ) arts, materials and function of the valves.

The licensee considers t11s an inconvenience, but not an unacceptable situation.

No violations or deviations were identified.

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6.

Review of Procedures and Safety Concerns for Transfer of Liquid Between Spent Fuel Pools (37702)

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For refueling of Unit 2 or 3, about 60,000 gallons of water must be transferred from storage tanks and between fuel pools using temporary systems.

The inspector reviewed the design and procedures to evaluate engineering adequacy and safety of these temporary systems.

The adequacy of the piaing and hoses with respect to the system pressure had been

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reviewed >y the resident inspector staff in 1987.

The ins)ector then had investigated the possibility that in the event of a hose areak or a seismicevent,thespentfuelpoolmayloosesignificantinventory.

The licensee determined that no damage would be ex)ected since the highest expected fluid velocity in the hose would be a)out 9 feet per second, and this along with the fact that the hoses are 3 inch noncollapsible flexible hose, showed that damage to safety related equipment was not likely.

In addition, the cuick disconnect couplings were not close to the areas of safety relatec equipment.

Also, concerning the possibility of loss of fuel pool inventory or flooding of safety related equipment as

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a result of a break, the high sump level alarm would be expected to alert o>erators, and after that{inued for the 30 minutes allowed for anDuring the low fuel pool alarm would sound.

t11s time, if leakage con operator to take action, the water in the basement of the auxiliary building would be expected to rise less than 3 inches above the floor.

The licensee determined that this would not be a threat to safety related equipment.

The licensee's evaluation appears to satisfactorily address the inspector's concerns.

No violations or deviations were identified.

7.

Implementation of Nuclear Safety Analysis Center (NSAC) 125 Guidelines (37702)

Because of the weakness in the safety evaluations associated with DCPs 6674 and 6621 discussed earlier, the inspector discussed weakness in safety evaluations with the Nuclear Safety Group (NSG).

The NSG is responsible for reviewing and evaluating licensee nuclear safety.

Based on discussions with NSG 3ersonnel, it was determined that the Nuclear Safety Analysis Center (iSAC) 125 Guidelines for Performance of Safety Evaluations would have helped to upgrade many of the safety evaluations

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NSG has reviewed over the last 6 months.

Although the licensee made internal commitments to implement these guidelines in October 1989, the en ineering procedures for performing safety analysis had not been im lemented as of the.date of this inspection.

Some of the NSAC 125 in tiatives have been implemented for Nonconformance Reports and training of both site and corporate engineers has been initia,ted for the licensee engineering staff.

After discussions with the inspector, the licensee made a commitment to I

implement the requirements of NSAC 125 for safety evaluations performed by the engineering staff by April 27, 1990.

The requirements are to be implemented with a revision to the procedure governing engineering evaluations.

No violations or deviations were identifie i.,

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Exit Meetino (30703)

The inspectors met with licensee representatives denoted in paragraph 1 on April 20 and May 4, 1990.

The scope and findings of the inspection were discussed as described in this report.

Licensee representatives acknowledged the inspector's findings.-

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