ML20195J724

Augmented Sys Review & Test Program Insp Rept 50-312/87-29 on 870928-1009.Open & Unresolved Items Noted.Major Areas Inspected:Open & Unresolved Items from Augmented Sys Review & Test Program Insp Rept 50-312/86-41

ML20195J724
Person / Time
Site:	Rancho Seco
Issue date:	01/04/1988
From:	Dyer J, Harper J, Haughney C, Howell A, Isom J, Norrholm L, Sharkey J NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD), Office of Nuclear Reactor Regulation
To:
Shared Package
ML20195J705	List: ML20195J700 ML20195J724
References
50-312-87-29, NUDOCS 8801280686
Download: ML20195J724 (41)

See also: IR 05000312/1987029

Text

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ENCLOSURE 2

U.S. NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

Division of Reactor Inspection and Safeguards

Report No.: 50-312/87-29

Licensee: Sacramento Municipal Utility District

P.O. Box 15830

Sacramento, California 95812

Docket No.: 50-312

Facility Name: Rancho Seco Nuclear Generating Station

Inspection Conducted: September 28, 1987 - October 9, 1987

Inspectors: 8thsW [4[8%

Date Signed

• J. E. Byer, Team Leader, NRR

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d4/ss

C. Har er, Metallurgical Engineer, NRR Date Signed

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A.i.jfowell,Rea@torOperationsEngineer,AEOD Dhte' Signed

,'Yb0 misn - /*b/ff7

(,*J. A. Fsom Reactor Operations Engineer, NRR Date Signed

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/,*J) M./Sharkey', @ctor Operations Engineer, NRl Date Signed

Consultants: *G. Morris, WESTEC; *D. Prevatte, WESTEC

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Accompanying Personnel: L. Miller, RV; R. Zimerman, RV; *D. Kirsch, RV;

• Crews, RV; *C. Myers, RV, *G. Perez, RV; ,

• A D'Angelo, RV; *D Baxter, EG&G; *M. Johnson, OED0 1

Reviewed By: i. M A /

L.' U Norrholm, Chief, Team Inspection Appraisal Date' Signed

and Development S tion #1, NRR

Approved By: v 1A _

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• C Haughney Chief 1 Inspection Branch, DttetSigned

• Attended Exit Meeting on October 9, 1987

8801280686 880125

PDR ADOCK 05000312

G PDR

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Scope:

An NRC headquarters team performed a special, announced inspectior to examine

the open and unresolved items from NRC's Augmented Systems Review and Test

Program (ASRTP) Inspection (50-312/86-41), assess the adequacy of the

licensee's expanded ASRTP (EASRTP) inspections, and evaluate the effectiveness

of the licensee's Engineering Action Plan (EAP) for improving the q'Jality of

engineering analyses. Additionally, the inspection team reviewed the

licensee's program for purchasing and controlling safety-related fasteners.

Results:

The inspection team closed 21 of the 44 open and unresolved itemt identified in

ASRTPinspection(50-312/86-41) dealing primarily with engineering, management,

and quality assurance areas. The remaining items concerning maiittenance,

operations, and surveillance testing will be reviewed during a fJture

inspection. The EAP appeared to improve the quality of design activities at

the station. The EASRTP inspections had been conducted in a manner comparable

with the NRC ASRTP inspection and identified significant safety issues. The

licensee's program for purchasing and controlling safety-related fasteners

appeared to have significant deficiencies; the NRC will review this program

again before restart. During this inspection, six new open items were

identified,

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TABLE OF CONTENTS

PAGE

1 INSPECTION OBJECTIVES .................................... 1

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2 STATUS OF PREVIOUSLY IDENTIFIED ITEMS .................... 2

2.1 Items Closed During the Inspection ....................... 2

2.2 Items Remaining Open After the Inspection ................ 9

3 DETAILED INSPECTION FINDINGS ............................. 17

3.1 Evaluation of Expanded ASRTP Inspection Program . .. . ... . .. 17

3.1.1 Revi ew of EASRTP Inspection Fi ndings . . . . . . . . . . . . . . . . . . . . . 17

3.1.2 Review of EASRTP Methodology ............................. 18

3.1.3 Comparative Inspection Results ........................... 20

3.2 Assessment of Engineering Action Plan .................... 21

3.2.1 Review of EAP Design Change Control Procedures ........... 22

3.2.2 R e v i ew o f EAP Wo r k P ro du c t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

3.2.3 EAP Independent System Reviews ... ....................... 25

3.3 Procurement and Control of Fasteners ..................... 26

3.3.1 Investigation of Counterfeit Fasteners ................... 26

3.3.2 Warehouse Material Controls .............................. 28

3.3.3 Purchase Order Review .................................... 28

3.3.4 Licensee-Identified Procurement Problems ................. 29

4 MANAGEMENT EXIT MEETING .................................. 31

Appendix A PERSONNEL CONTACTED

Appendix B DOCUMENTS REVIEWED

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1 INSPECTION OBJECTIVES

The objectives of this inspection were to examine the corrective actions

taken by the Sacramento Municipal Utilities District (SMUD) as a result of the J

NRC's Augmented System Review and Test Program (ASRTP) inspection conducted at

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Rancho Seco Nuclear Generating Station in February 1987. This effort included:

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(1) a review of the specific open and unresolved items identified during the

ASRTP inspection; (2) an evaluation of the licensee's Engineering Action Plan

developed to improve the quality of ongoing engineering analyses and calcula-

tions, and (3) an assessment of the licensee's completed pnrtions of the

Expanded ASRTP (EASRTP) inspections performed on 33 safety-related systems to

better ensure safe design and operation. Additionally, the team reviewed

SMUD's program for procuring and controlling safety-related fasteners.

2 STATUS OF PREVIOUSLY IDENTIFIED ITEMS

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ihe inspection team reviewed the status of the open and unresolved items

identified during its ASRTP inspection (50-312/86-41). In general, those items

associated with engineering, quality assurance, and management areas were

closed, but items concerning maintenance, operations, and surveillance testing

remained open. These open items will be reviewed during future inspections.

2.1 Items Closed During the Inspection

Open Item 50-312/86-41-02: Auxilibry Feedwater Flow Above the Maximum

Design Rate

Problem 54 of the AFW (auxiliary feedwater) System Status Report (SSR) stated

that AFW flow to a once-through steam generator (OTSG) could exceed the maximum

design rate of 1800 gpm. The team disagreed with the licensee's plans to

correct this problem after restart.

The licensee installed a cavitating venturi in each OSTG line to limit flow to

1000 gpm. The team reviewed the design change package and found no deficien-

cies. This item is closed.

Open Item 50-312/86-41-03: AFW Full Flow Test Line Instrumentation

Problem 3 of the AFW SSR stated that the instrumentation on the AFW full-flow

test line was inaccurate because downstream piping was subjected to condenser

vacuum. This deficiency contributed to further problems with measuring AFW

pump capacity using alternate methods as described in unresolved item

50-312/86-41-28. At the ASRTP inspection exit meeting, the licensee committed

to correct this problem before restart.

The licensee installed a restricting orifice downstream of the flow element to

increase back pressure. The element was resized accordingly and sensing lines

were redesigned to eliminate air entrapment. The team reviewed the design

change packages for these modifications and found no deficiencies. This item

is closed.

Open Item 50-312/86-41-04: AFW Pump Runout Damage

Problem 43 of the AFW SSR outlined the resolution for verifying that the AFW

pumps had not been danaged by the pump runout condition that occurred during

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the December 26, 1985 transient. The licensee had not planned to contact the

vendor to verify that the internal components of the pump had not been damaged.

The licensee subsequently contacted the pump marufacturer who identified which

internal components should be checked or replaced. The licensee completed the

necessary work. This item is closed.

Open Item 50-312/86-41-05: AFW Pump Performance Calculation

Calculation Z-FWS-M2081, performed in response to AFW SSR Problem 55, was found

to have an error that would have identified an incorrect and nonconservative

acceptance value for pump performance testing.

The licensee replaced this calculation with Z-FWS-H2237. The team reviewed this

calculation and found no discrepancies. This item is closed.

Unresolved Item 50-312/86-41-06: Instrument Air System Design For AFW

Flow Control Valves

Several deficiencies were identified in a proposed modification to install

backup instrument air bottles to supply air to AFW flow control valves. The

licensee's resolution for these items is as follows:

(1) Inadequate Valve Actuator Overpressure Protection: The pressure relief

setpoint was reduced to less than the design pressure of the main and

startup feedwater contrt' valve actuator diaphragms. Additionally, an

airset was installed in v a air supply to the diaphragms providing addi-

tional protection.

(2) Lack of Seismic Qualification of Air Valves: The licensee performed

calculation Z-ZZZ-C0863 to confirm the seismic qualification of the

control valves, excess flow valves and adjustable check valves for the

backup air system.

(3) Incorrect Check Valve Design: The excess flow check valves were replaced

with adjustable check valves which were better suited to the design

application in the backup air system.

(4) Monitoring of Backup Air Supply Pressure: Operating logs were issued

requiring the recording of bottle pressures on a daily basis. However,

the acceptable pressures identified on the log sheets were at the 2 and

3-hour alarm points for the bottles. These pressures were well below the

minimum specified by Engineering in the D3 sign Basis Report for the

modifications. The licensee initiated a revision to the operating logs to

establish new minimum bottle pressures which will accomplish the intended

purpose.

(5) Backup Air Supply Test Procedure: The licensee developed a periodic test

procedure for the backup air system. At the time of the inspection, the

procedure was being reviewed for approva*.

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(6) Incorrect Design of the Pressure Control Valve: The valve manufacturer

certified that the valve would give the tight zero demand performance

desired for the system. The combinatior, of the manufacturer's statement,

the proposed testing, and the daily logging of pressure convinced the

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inspection team that reasonable assurance of pressure control valve

performance will be provided.

(7) Incorrect Fabrication Drawings: The drawings were changed to show the

proper orientation of the valves.

The team concluded that the above actions were appropriate. This item is closed.

Open Item 50-312/86-41-08: Environmental Qualification of Emergency Feedwater i

Initiation and Control (EFIC) System Excess Flow Valves. 1

The excess flow valves used in the instrument lines for OSTG 1evel instruments  !

installed as a part of the EFIC system modifications were not environmentally  !

qualified. I

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The licensee's analysis of calculation Z-EQP-E0689 qualified the critical

components (0-rings) for 10 years. However, the basis for the acceptance

criteria for thermal aging in a dynamic application was not included in the

analysis, and the 10 year replacement requirement for the 0-rings had not been

included into the appropriate maintenance procedures. During the inspection,

the licensee revised the calculation to include the basis for the acceptance

criteria and inserted the requirement for replacement of the 0-rings into the

appropriate maintenance procedures. This item is closed.

Unresolved Item 86-41-09: EFIC System Single Failure Susceptibility

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The EFIC system sensing instrumentation was designed so that certain failures

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would activate unnecessary protective systems. This susceptibility appeared to

be contrary to the updated safety analysis report and had not been reported to

the NRC as is required by 10 CFR 50.59.

The licensee revised the safety analysis for the EFIC system modification to

include responses that could be expected with spurious initiation. None of the

responses were outside the design bases of the plant. This item is closed for

the inspection report purposes. The NRC is currently reviewing the adequacy of

this design as part of the safety evaluation report (SER) for plant startup.

Open Item 50-312/86-41-10: Maintenance Bypass Testing of the EFIC System

The proposed test for the EFIC system did not verify that whenever any EFIC

channel was placed in the maintenance bypass position, the remaining channels

would not be inhibited or degraded.

The inspection team reviewed the revised special test procedure (STP) 666,

"EFIC Cold Functional Test," and determined that the maintenance bypass feature

was adequately tested. This item is closed.

Unresolved Item 50-312/86-41-15: Condensate Storage Tank Pressure Relief and

Vacuum Protection

The setpoint tolerances for the condensate storage tank (CST) pressure relief

valves were such that the valves could be set above the design pressure for the

tank (setpoint as high as 2.5 psig vs. 2.0 psig design pressure).

Additionally, the CST was designed for a 1.0-inch water vacuum, but a

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calculation in the file indicated that this condition could be exceeded without

assuming a single-valve failure.

The licensee changed the relief valve setpoint tolerances to ensure that the

design pressure would not be exceeded. Additionally, an analysis (Z-MCM-M2212)

was performed that showed the tank capable of withr, Landing 5.0 psig pressure

and 1.74 inches water vacuum. This analysis also snowed that for the maximum

outflow with one vacuum 'reaker failing to open, thi!, new design value for

vacuum would not be exceeded. This item is closed.

Unresolved Item 50-312/86-41-16: DC System Short-Circuit Calculations

Inconsistencies were identified with the calculation of the battery's contri-

bution to a short circuit. This matter included inconsistencies in the

manufacturer's data referenced in, and attached to, :he calculation. The

inspection team's rough calculations indicated that ;ome circuit breakers may

not have been properly sized for the de system.

The licensee replaced existing de cables with smallec cables to reduce short-

circuit current and revised the calculation (Z-DCS-E)612) to incorporate the

team's comments. Additionally, the team also verified that the design inputs i

for the smaller cable, added to limit the short circuit current, were correctly

factored back into the new voltige calculations. This item is closed.

Unresolved Item 50-312/86-41-17: Battery Sizing Calculation

The staff was concerned about the load profile and ninimum design temperature

used for sizing the new batteries installed in the ruclear services electrical

building (NSEB).

The licensee revised the calculation (Z-DCS-E0636) t.o respond to the team's l

concerns. The team reviewed this revised calculation and found no deficien-

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cies. This item is closed.

Open Item 50-312/86-41-18: Drawing E-101 Error

A discrepancy between calculation Z-EDS-E0076 and the main single line diagram

E-101 was identified. The team confirmed that the drawing was wrong, not the

calculation.

The licensee issued a drawing change notice (DCN) to correct the drawing. The

DCN was reviewed by the team for the identified concerns and found acceptable.

This item is closed.

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Unresolved Item 50-312/86-41-19: AC Short Circuit Calculatio_n '

An apparent discrepancy was discovered between the startup transformer

impedance given on the unit's nameplate and that derived from the unit's test

data. The licensee could not immediately confirm which value was correct.

The licensee revised its calculation (Z-EDS-0120) using the more conservative

value and demonstrated that the design was adequar.e. The discrepancy between

the nameplate and test data impedances wa,s resolved by the vendor's statement

that the European nameplate information contains design data only. This item

is closed.

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Unresolved Item 50-312/86-41-20: Battery Charger Cable Size

The ampacity of the d: power cable from battery charger H4BAC to de bus SOC

would not meet the commitment to size power cables at 125 percent of the full

load current that was made in the "Rancho Seco Nuclear Generating Station

Updated Safety Analysis Report" (USAR).

The licensee respondtd with calculation Z-EDS-E0744 and proposed administrative

procedure changes that would justify operating this particular cable in its

overload region for a limited number of times over the life of the plant. The

response sufficiently justified the overload condition of the cable for this

specific cable. The response did not address the USAR commitment to size cable

at 125 percent of the full load current. However, because this cable is in a

dedicated conduit and its use will be administratively controlled, the team

found the ampacity of cable acceptable for limited use. This item is closed.

Unresolved Item 50-312/86-41-21: IDADS Alarm Interface Problems

The required time delay on pump startup for the AFW pump runout alarm had not

been provided either by hardware or software design. Additionally, the team

found that the voltage for the NSEB de system buses was being monitored incor-

rectly by the Interim Data Acquisition and Display System (IDADS) as digital

inputs instead of aralog inputs.

The licensee made scftware changes to the IDADS alarn for the AFW pump runout

alarm and issued Engineering Change Notice (ECN) R151 to modify the inptt

circuit hardware for de bus voltage. The team reviewed these modifications and

found no discrepancies. This item is closed.

Open Item 50-312/86-41-22: Motor Operated Valves Overload Protection

No thermal overload protection or overload alarms existed for the safety-

related motor opera;ed valves (MOVs). The licensee's response during the

initial inspection was that it was more important that a safety-related MOV

fail attempting to perform its safety function than to trip because of an

erroneous response from an overload relay. No further work was performed in

this area because tie licensee planned to resolve this problem after restart.

The team agrees that this was not a restart item that needed immediate atten-

tion. The team was concerned, however, about undetected long-term degradation

of the motor's insulation and the resultant increased probability for valve

failure. This iter is closed.

Open Item 50-312/86-41-23: Safety Evaluations

The licensee's procedures for conducting safety evaluations in accordance with

10 CFR 50.59 lacked sufficient guidance for conducting the evaluations; also

there were no qualification requirements for personnel performing the

evaluations.

The licensee issued revised procedure RSAP-0901, "Safety Review of Proposed

Changes, Tests, and Experiments," for conducting safety-evaluation required by

10 CFR 50.59 and implemented a training program to qualify personnel performing

the evaluations. This item is closed.

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Unresolved Item 50-312/86-41-24: Drawing Control

Deficiencies were identified with the procedure for and implementation of a )

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drawing control program.

The licensee developed and implemented administrative procedures RSAP-0503,

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"Design Change Document Control," and RSAP-0505, "Site Document Control (SDC)

Distribution Control," for drawing control. Additionally, the licensee {

conducted a thorough audit of their drawing control program to verify imple- l

mentation of the new procedures. The team verified that the specific deficien-

cies identified in inspection report 50-312/86-41 were corrected.

Also, a random sample of approxinately 20 controlled drawings were

reviewed in four locations, with no noted discrepancies. This item is closed.

Unresolved Item 50-312/86-41-25: Engineering Calculations

Several programmatic deficiencies were found with calculatinns performed by the

licensee in support of system design. As a result, the licen;ee implemented an

Engineering Action Plan (EAP) to improve all calculations. The EAP evaluation

is discussed in Section 3.2 of this report. This item is closed.

Unresolved Item 50-312/86-41-26: Expired Inservice Test (IST) Program

The first 10-year Inservice Test (IST) Program that the licensee submitted had

apparently expired on April 6, 1986 and no extension or new program had been

submitted to the NRC for approval.

On June 25, 1987, the licensee submitted its second 10-year IST Program to the l

NRC for approval. The NRC is currently reviewing this program. This item is l'

closed.

Unresolved Item 50-312/ 86-41-27: Pump and Valve Test Data Trending Program

The licensee was not entering test data into the Inservice Inspection Log used i'

for trending test data, although testing was conducted.

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The inspection team reviewed the licensee's new program for trending IST data

which included both a data log and a graphic trending program. These appeared

adequate. Additionally, the licensee is planning to implement a computer aided

trending program after restart. This item is closed.

Open Item 50-312/86-41-29: Inconsistent Stroke Times for AFW Flow Control

Valves

Identical AFW flow control valves FV-20527 and FV-20528 had inconsistent stroke

time acceptance values. There was no technical justification for the different

times since these valves appeared to be identical valves in identical

applications.

The licensee verified that the acceptance values should be the same. A special

test procedure was being developed to establish consistent acceptance values

during AFW functional testing. This item is closed.

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Unresolved Item 50-312/86-41-32: Systems and Equipment

Numerous examples of administrative deficiencies were found involving the safe

clearance tag system, the abnormal tag system, and the Information Sticker Log.

It appeared that operations personnel did not always pay sufficient attention

to detail to ensure adequate control of plant and system status.

The licensee issued Special Order 87-19, "Operations Audits," that discusses

the Operations Department requirements for auditing each of the above

referenced tagging systems. The team reviewed the administrative control

systems and verified that the previously identified weaknesses were corrected.

The team conducted a detailed evaluation of the Information Sticker Log and the

Abnormal Tag Log and noted no deficiencies. Additionally, the licensee had

recently implemented Procedure AP.90, "Work and Test Authorization (WATA)

Program " Revision 1, that documented the authorization of work and testing

that affected the operational status of components and system; tracked the

operational status of components and systems; and prevented removing a

Technical Specification required system or component from service while the

redundant train was inoperable. The team concluded that the implementation of

the WATA program will significantly improve the ability of the Operations

Department to effectively control and track equipment status. This item is

closed.

Open Item 50-312/86-41-35: Operator Training Materials

System training manuals were neither controlled nor maintained current by the

Training Department. Uncontrolled copies of these training manuals were found

in the control room.

The licensee placed an index manual with the four volumes of system training

manuals in the control room. This index featured a list of working and c>m-

pleted ECNs sorted by system. This list alerted the user to other documents

that could modify the system. Additionally, the Operations Department it, sued

procedure AP-23.06, "Operation Procedures," Revision 0, which listed spe:ific

station procedures that were approved for use and prohibited the use of system

training manuals in lieu of approved procedures. Finally, the licensee planned

to revise the system training manuals in 1988.

The ASRTP inspection team was also concerned that several operating and

casualty procedure revisions were too complex to be adequately addrersed by the

Operator Reading Assignment Program.

The licensee improved its scheduled training for the new procedurer,. The team

reviewed the procedure revisions and the Training Department's methods for

selecting procedures for training. On the basis of this review, the team

concluded that the degree of procedure training fer each new pro:edure could be

appropriately addressed by the proposed classroom training, simulator training,

procedure walkthrough, or the Operator Reading Assignment Program. The inspec-

tion team did not review the technical adequacy of the planned training for the

new procedures. This item is closed.

Open Item 50-312/86-41-37: Maintenance Trending Program

The licensee had not implemented its maintenance trending program as described

in procedure AP-650, "Preventive Maintenance Program," Revision 5.

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Maintenance Administrative Procedure MAP-009, "Preventive Maintenance Program,"

was issued and superseded AP.650. The licensee subsequently implemented the

maintenance trending program described in MAP-009 that included, in part, a

vibration monitoring program for rotating equipment. Although only recently

implemented, there was objective evidence that the program was effectively

administered. For example, the licensee had projected the iminent failure of  ;

a service water pump (P-9768) thrust bearing on the basis of the results of

routine vibration testing. The licensee was able to replace the pump bearing

before a catastrophic failure occurred. The licensee planned to implement the

use of trending techniques in other areas of the predictive maintenance

programs using recently acquired computer software programs. This item is

closed.

Open Item 50-312/86-41-38: Quality Assurance (QA) Audit Program Deficiencies

Significant deficiencies were found with the scheduling, performance,

responses, followup, and Management Safety Review Committee (MSRC) oversight of

the QA audit program.

The licensee reorganized the quality assurance (QA) organization, providing new

emphasis and direction to the audit program. The team reviewed the improved QA

audit program and found that the licensee currently had no audits overdue. In

addition, a requirement was added to Procedure QAIP-1, "Quality Assurance

Audit Procedure" to issue audits within 30 days after the inspections were com-

pleted. All audit findings were addressed to the department manager of the

area reviewed and a response was required within 30 days after the report was

issued. Also, at the time the responsible department replied to the audit '

finding, a completion date was submitted for QA concurrence. It was the job of

the QA organization to verify that the responsible organization had taken the

appropriate corrective action. The audit finding was then tracked using the

Management Sumary Report that highlights both overdue audit responses and

corrective actions. This report was distributed monthly to the Chief Executive

Officer, Nuclear and to both Assistant General Managers. Lastly, the licensee

had formed an MSRC Quality Oversight Subcommittee to advise the MSRC on matters

related to the QA program. Specifically, the subcomittee reviewed audit

program findings and closure reports to ensure the effectiveness and timeliness

of the QA program. The subcomittee was required to report the results of its

review to the MSRC. This item is closed.

Open Item 50-312/86-41-40: Corrective Action Programs

Deficiencies were identified with the trending program for nonconformances, J

numerous outstanding NRC open or unresolved items, and lack of a program that l

allowed significant conditions adverse to quality to be brought to the atten-

tion of the appropriate level of management.

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The licensee implemented procedure QAIP-16. "Trend Analysis Program," Revision

0, which related deficiencies to a common cause, identified recurring problems,

and defined the extent and severity of the problems. Additionally, the licen-

see reviewed and closed out numerous NRC open items and is continuing to do so.

The licensee issued procedure QAIP-27, "Corrective Action," Revision 0, which

allowed the QA organization to elevate issues of a significant nature to the

appropriate level of management. This item is closed.

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Open Item 50-312/86-41-41: LRS Management Appraisal Report Issues

The licensee did not include outstanding open items from the LRS consultants'

Management Appraisal Report in the ongoing Plant Performance and Management

ImprovementProgram(PP&MIP).

The licensee added the open issues from the LRS report in its new tracking

program as described in procedure RSAP-0215, "Restart Scope List (RSL)/Long

Range Scope List (LRSL) Development and Administration," Revistor. O. The

RSL/LRSL superseded the Plant Performance and Management Improvement Program

(PP&MIP). This item is closed.

Open Item 50-312/86-41-42: Validation of the 00I-12 Process

During the ASRTP inspection, the licensee committed to validate the adequacy of

the reviews conducted as part of the QCI-12 review process.

The ifcensee implemented the EASRTP inspection program for all 33 selected

systems. The EASRTP inspection program appeared to contribute to improving

safety at the plant as discussed more fully in Section 3.1 of this report.

This item is closed.

Open Item 50-312/86-41-43: Selected System Status Report (SSR) Document

Control

The licensee did not appear to be properly controlling SSRs considering their

use as a basis for the NRC Safety Evaluation Report (SER) for restart authori-

zation.

The licensee issued procedure AP-93, "System Status and Investigation Reports,"

Revision 0, which adequately described the control procedures for the SSRs.

This item is closed.

Open Item 50-312/86-41-44: Restart Organization

Because the licensee's ' organization had several vacancies, temporary contractor

personnel filled key positions in the restart organization. Additionally,

there were no plans for making the transition from contractor personnel to

permanent employees to support restart.

The licensee has reorganized and contracter personnel were replaced in key

man 6gement positions with permanent SMUD employees. The inspection team

reviewed the qualifications of management' personnel and interviewed selected

managers. New personnel filling key positions appeared to be well qualified I

and were working toward becoming an effective management team. This item is

closed.

2.2 Items Remaining Open After the Inspection

Open Item 50-312/86-41-01: AFW Turbine Driven Overspeed Issues

The team identified the following concerns associated with overspeed of the

turbine-driven AFW pump: (1) The overspeed trip point (as high as 4650 rpm) I

could be well above the overspeed rating of the electric-drive motor (4320 rpm)  !

connected to the comon shaft, (2) the pump discharge piping was not analyzed  !

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for the overpressure condition that could result from an overspeed event, and

(3) the time required for the depressurization of the turbine governor control

oil after overspeed trip was not known but was required for operating procedure

guidance to prevent a subsequent overspeed trip upon restart.

The licensee obtained a certification from the motor vendor that the motor was

not overstrested up to 4500 RPM and the tolerance on the turbine overspeed

setting was changed so as not to exceed 4500 RPM. An analysis was perfonned of

the discharge piping. All but one section was found to be within allowable

stresses and that one section was only slightly above the allowable strasses

for worst case conditions. The licensee committed to replace this section of

piping before restart. An overspeed test is planned to detennine the governor

bleed time. The test results will be incorporated into the operating procedure

to prevent overs ed on subsequent restarting of the AFW pump. The team found

the licensee's actions in response to this item acceptable. This item will

remain open until the results of pump testing are available and procedures are

revised to reflect those results.

Open Item 50-312/86-41-07: Main Feedwater (MFW) System Problems

The inspection team identified eight SSR problems that should be resolved

before restart. The status of each of those problems follows.

Problem 6: Faulty Main Feedwater Pump Lovejoy Control Response

.

The licensee developed a permanent modification to jumper out the dead

band module. The licensee determined that this configuration would be

more reliable. The inspection team found this response adequate.

Problem 9: Main Feedwater Startup Flow Control Valves Stick Closed

Occasionally and have Slow Response to OTSG Level Change

The licensee refurbished the flow control valves to correct the observed

problems during this outage. The team found the licensee's response to i

this problem adequate.

Problem 12: Main Feedwater Flow Control Valve Positioning During

Transients (Overfeed)

The inspection team reviewed Casualty Procedure C.10 "Main Feedwater

Induced Transients," Revision 2, and found that the prescribed guidance

adequately addresses actions for an overfeed condition.

Problem 18: Correct Casualty Procedure C.26 for Main Feedwater

Pump Operation With Low Condenser Vacuum

This problem will be reviewed during a future inspection.

Problem 19: Correct Casualty Procedure C.10 for Action on Lo,sss

of One Main Feedwater Pump

The inspection team reviewed procedure C.10, "Main Feedwater Induced

Transients," Revision 2, and found the guidance for the loss of one main

feed pump to be adequate.

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Problem 32: Update Piping and Instrumentation Drawing M580, Sheet 1

The team reviewed the revised drawing and found no discrepancies. The

inspection team found the licensee's response to this problem adequate.

Problem 31: Main Feedwater Pump Control From Lovejoy to Bailey Hand / Auto

Station Is Not Performing as Designed

The licensee plans to revise Procedure A.50, "Main Feedwater System," to

ensure that controller inputs are matched before the shift occurs from

manual to automatic. The inspection team will review procedure A.50 at a

future inspection to determine whether the problem has been resolved.

Problem 45: Main Feedwater Pump Governor Is Slow to Respond

The dead band module was found to be the failing component. A modifica-

tion was generated to remove the dead band module and install a jumper.

This, in effect, was the same modification as the temporary jumper that

had been installed during previous plant operations and found to be

satisfactory. This change required the controller to be continually in

service rather than being in service only when the pump is operating at a

speed that is outside the dead band. The licensee determined that this

configuration was more reliable. The team found this response to be

adequate.

Open item 86-41-07 will remain open pending the closeout of MFW SSR problems 18

and 31.

Open Item 50-312/86-41-11: 125 V de System Problems

The inspection team identified four SSR problems that should be resolved before

restart. The status of each of those problems follows:

Problem 8: Operating Procedure A.61 Deficiencies. The revised procedure

will be reviewed during a future inspection.

Problem 9: Battery Room Temperature Control. Discrepancies existed

between the temperatures assumed in the battery sizing calculations and

the referenced mechanical heating, ventilation, and air conditioning

(HVAC) calculations. The licensee issued new calculations to establish

the allowable temperatures for the auxiliary building and NSEB battery

rooms. The licensee's electrical group prepared new battery sizing

calculations that had minimum battery temperatures traceable to the new

HVAC calculations. Additionally, maintenance procedures were revised

accordingly to limit the battery minimum and maximum temperatures to the

design input valves. The inspection team found this response adequate.

Problem 20: Battery Charger Refurbishment. The licensee had identified

one of six original Class lE battery chargers for refurbishment before

restart, although the filter capacitor banks for all six chargers were

overdue for replacement. The licensee will replace the filter capacitor

banks for all six chargers before restart. The inspection team found this

response adequate.

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Problem 22: Damaged Battery Terminal Posts. The licensee prepared

nonconformance reports (NCRs) 55504 and 55548 to identify the deformed

terminal posts on two vital battery cells. The manufacturer's representa-

tive inspected the cells on-site and concluded that the damage to the

posts would not affect the safety function of the batteries. The licensee

planned no further action. The inspection team found this response 1

adequate. l

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Open item 86-41-11 will remain open pending resolution of SSR problem 8. '

Open Item 50-312/86-41-12: 120-V ac System Problems

The inspection team identified five SSR problems that required further resolu-

tion before restart. The status of each of those problems follows:

Problem 2: Control Room Indication for Breakers. The licensee had i

identified a number of problems with inadequate control room indication of I

circuit breaker position in the ac vital electrical systems. The licensee

contracted to have Impell review the vital ac systems to determine the

adequacy and consistency of the alarms resulting from a trip of any l

electrical protective device. The summary report identified 15 generic

recommendations requiring both procedure and hardware changes. The inspec-

tion team reviewed the proposed changes and determined that the procedure

changes would provide adequate guidance for restart. The licensee j

initiated the required procedure changes. The inspection team found this

response adequate.

Problem 5: 120 V ac Systems Casualty Procedures. This item will be

reviewed during a future inspection.  :

Problem 6: Missing Load Schedule in Procedure A.62. The licensee revised 1

Procedure A.62, "120 Vac Vital System," to include the missing load list.

The inspection team found this response adequate.

Problem 7: Response to IE Bulletin 79-27. NUREG-1195 identified a number

of problems involving an inadequate response to IE Bulletin 79-27, "Loss

of Non-Class 1E Instrumentation and Control Power System Bus During Opera-

tion." The licensee had planned to review and update its original IE

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Bulletin response, but, before the original ASRTP inspection, had not

planned to impleme-t the recommendations resulting from the review before

restart. The licensee c.ontracted with Babcock & Wilcox (B&W) to review

the requirements of IE Bulletin 79-27. The B&W report listed 13 recom-

mendations including procedure changes and provisions for alternate power

supplies. These recommendations were incorporated into the licensee's

response for IE Bulletin 79-27. The inspection team reviewed the response

associated with the Class 1E control power systems and determined that

one recommendation was significant enough to require a hardware modifica-

tion before restart. A spare 120 Vac cable should be dedicated to provide

an alternate power source to the electrically actuated pressurizer relief

valve (PSV-21511). The NRC is reviewing the licensee's complete response

to IE Bulletin 79-27 and will include the results of that review in a

supplement to the SER for re; tart.

Problem 12: Local Indication of Circuit Breakers Position. The licensee

identified a number of problems with inadequate local indication of

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tripped circuit breakers. An Impell review indicated that this problem

was limited to Westinghouse circuit breakers. The team determined that

Westinghouse circuit breakers in panels S1A3, S1A4, S183, and S184 had

trip indicators that may be questionable. The licensee agreed that these

circuit breakers should be improved by adding a white dot to their handles

to help the operator detect a tripped circuit. The implementation of

this resolution will be reviewed during a future inspection.

Open item 86-41-12 will remain open pending resolution of 120-V ac SSR problems

5, 7, and 12.

Open Item 50-312/86-41-13: 480-V ac System Problems. The inspection team

identified nine SSR problems which required further resolution before restart.

The following is a status of those problems:

Problem 10: Loss of Power Alarms for 480 Vac MCC Loads. The licensee

questioned whether acceptable alarm or indication was available to detect

loss of power (by tripping the circuit protective device) to the 480 Vac

Motor Control Center (MCC) loads. An Impell study identified acceptable

alternate indication for all but three MCC loads. The team determined

that the three identified problem circuits (Breakers 2A317, 28317, and

2A208A) should have their circuit breaker handles modified to provide

clear indication of circuit breaker trip status for operators (See

problem 16).

Problem 11: 480-V ac System Casualty Procedures. This item will be

reviewed during a future inspection.

Problem 16: 480-V ac Circuit Breaker Handle Position. The licensee

identified a problem with the inability to visually determine when a motor

control center circuit breaker had tripped. The MCC manufacturer

(Westinghouse) made recommendations for modifying the circuit breaker

handles to give adequate visual indication of the state of the breaker

(closed or tripped). Two ECNs were prepared to perform this work. The

licensee determined that the three circuits identified in SSR Problem 10

as having no other means of identification should have their circuit

breaker handles modified before restart. The remaining circuit breakers

will be modified after restart. The inspection team found this response

adequate.

Problem 19: Inconsistent Alarms for loss of 480-V ac Systems. The I

licensee identified an inconsistency in the manner of alarming the feeder

and incoming breaker trips on the 480-V ac switchgear. The licensee ,

provided a point paper stating that the depth of the inconsistency was of I

minor concern. The team agreed that this minor inconsistency should not

be a restart item and found this response adequate.

.

Problems 25, 33, and 34: Procedure A.59 "480 Vac System Operating Procedure"

Deficiencies. This item will be reviewed during a future inspection.

Problem 26: Indication for loss of Control Power for 480-V ac Circuit

Breakers. The licensee questioned the potential lack of indication for

loss of 480 Vai switchgear control power. A study performed by Impell

indicated that all 480-V switchgear circuits contain one or more indicat-

ing lights that will turn off when control circuit power is lost. The

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team reviewed random examples of 480-V switchgear circuits, observing the

referenced indicating lights on the control circuit schematics drawings,

and found no deficiencies. The team found this response adequate.

Problem 46: Drawing E-108, Sheet 30 Deficiencies. The licensee identi-

fied a potential conflict between the single line diagram for bus S2C9 and

the schematic drawing for breaker 2C915. On further review, the licensee

noted that both drawings were consistent and that breaker 2C915 was a

spare. Bus SIGB was a non-safety instrument bus fed from a battery-backed

inverter and did not require a backup power source. The inspection team '

agreed with this review and found the response adequate.

Open item 86-41-13 will remain open pending resolution of 480-V ac SSR problems

25, 33, and 34.

Open Item 50-312/86-44-14: 4160-V ac System Problems.

The inspection team identified four SSR problems that had to be resolved before

restart. The status of those problems follows.

Problem 8: Slotted Protective Relays. The licensee found that the

mounting holes for magnets used in the protective overcurrent relays

mounted on the 4160-V switchgear had been elongated and the seismic

effects of the modification had not been adequately evaluated. The

licensee performed a calculation (Z-EDS-C0929) to prove the seismic

acceptability of the magnet mountings. The inspection team found this

response adequate.

Problem 25: Indication of Loss of Control Power for 4160-V Switchgear.

The licensee identified a potential problem involving indication of loss

of control power to the 4160-V switchgear control circuits. An Impell

study on this subject identified five circuits that could lose control

power and not be detected. The team determined that the indicating lights

for the control circuits associated with AFW pump P-319 and the under-

voltage trip circuit logic associated with the four safety-related 4160-V

buses should be modified to provide direct indication in accordance with IE

Bulletin 79-27.

Problem 32: Procedure for Operating the Startup Transformer. The

licensee found that no guidance was provided for operating the startup

transformer. The team reviewed the revised procedure A.54, "220 KV

Electrical Systems," Revision 9, and found that the addition of the

startup transformer system checks provided were adequate guidance.

Problem 33: Casualty Procedure for Loss of Cooling to Transformer.

Casualty procedure C.143, "Loss of 480 Volt MCC S2El," incorrectly

directed the operator to check the alternate power supplies on the loss of

power to both auxiliary transformers, ACB-2E148 and ACB-2E149, and both

startup transformers, ACB-2E150 and ACB-20151. Casualty procedure C.143 I

was revised to direct the operator to check the normal power supplies to I

these transformers upon loss of power. The team found this response l

adequate.

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Open item 86-41-14 will remain open pending resolution of 4160-V ac SSR Problem I

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Unresolved Item 50-312/86-41-28: AFW Pump Operaoility

The team was concerned that AFW flow testing did not reveal the actual AFW flow l

to the OTSGs because there was no valid basis for the 60 gpm minimum flow j

bypass flow rate that was used in determining the flow available to the OSTG's -

and the CST level instrument used to determine the total flow out of the CST i

was found to be unreliable and not properly calibrated. '

With the installation of the modifications described in the response to ASRTP

item 86-41-03 to allow accurate measurement of pump flow directly tt. rough the

full-flow test line, the concerns listed above are no longer pertinent to

determining the flow of the AFW pumps. A test procedure was developed to

determine the flow in the minimum-flow test line during operating conditions.

The licensee has also committed to performing a multi point calibration of the

CST level instrument LI-35803. This item will remain open pending completion

of the minimum-flow test procedure and performance of a full-loop, multi point

calioration of the CST level instrument LI-35803.

Unresolved Item 50-312/86-41-30: AFW System Surveillance Procedures

This item will be reviewed during a future inspection.

.

Unresolved Item 50-312/86-41-31: 125-V de System Surveillance Testing

Deficiencies were identified with the observed cell temperatures and electro-

lyte levels in the station batteries and recorded values for the battery

charger periodic maintenance tests. These deficiencies were caused by

inadequate procedures.

The licensee revised procedures SP.300, "Weekly Nuclear Service Battery Pilot

Cell Test," and SP.301, "Monthly Nuclear Service Battery Test." The inspection

team reviewed the revised procedures for the battery testing and found them

acceptable. The revised procedures for battery charger testing will be

reviewed during a future inspection. This item will remain open. '

Open Item 50-312/86-41-33: AFW Operating Procedure Deficiencies

This item will be reviewed during a future inspection.

Open Item 50-312/86-41-34: AFW Pump Runout Recognition and Control

This item will be reviewed during a future inspection. l

Unresolved Item 50-312/86-41-36: Valve Maintenance Procedures

The licensee's maintenance procedures for MOVs provided inadequate guidance for

inspection of the brakes and the maintenance procedures for safety-related

air-operated valves were not developed.

The licensee's corrective action for instituting maintenance procedures for

MOVs with brakes and safety-related, air-operated valves were reviewed. The

applicable maintenance procedures for the MOVs were revised to include the

vendor recommendations for inspecting the brakes. The condition of the brakes

will be examined every refueling cycle. The licensee's corrective action for i

establishing maintenance procedures for safety-related, air-operated valves had l

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not been completed at the time of this inspection. This item will remain open

pending verification of the development and implementation of maintenance

procedures for safety-related air operated valves.

Open Item 50-312/86-41-39: QA Surveillance Program

The team found (1) deficiencies in the quality of past QA surveillances, (2)

lack of accountability for correcting identified deficiencies, and (3) lack of

a trending program.

The licensee revised procedure QAIP-2, "Quality Assurance Surveillance

Program," to require that responsible organizations respond to findings within

14 days of receiving a report. Additionally, procedure QAIP-16, "Trend

Analysis," has been developed to trend results of the QA Surveillance Program

with other inputs. These changes were only recently issued and had not been

implemented sufficiently for the inspectors to adequately review the improved QA

Surveillance Program. This item will remain open.

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3 DETAILED INSPECTION FINDINGS

3.1 Evaluation of Expanded ASRTP Inspection Program

In response to the NRC Augmented System Review and Test Program (ASRTP)

concerns about the adequacy of the engineering and technical reviews performed

as part of the SRTP process, the licensee decided to perform expanded ASRTP

(EASRTP) inspections on all 33 selected systems using 6 multi-discipline

teams. The EASRTP inspections were patterned after the NRC ASRTP inspection

(50-312/86-14) and were intended to ensure that pertinent areas of system

operation, design, testing, and maintenance were reviewed before plant restart.

The NRC inspection team conducted its evaluation of the EASRTP inspections in

three ways:

(1) Reviewed issued EASRTP inspection reports for significant findings.

(2) Reviewed inspection documentation to detennine thoroughness of

reviews.

(3) Conducted independent reviews of the same systems to compare inspec-

tion findings.

At the time of the inspection, the EASRTP inspection team had completed

its investigation of 18 selected systems and was progressing on schedule

to complete inspection of the 33 systems by November 1987.

3.1.1 EASRTP Inspection Findings Review

After completing 18 systems inspections, the EASRTP team had identified more

than 150 findings that had been accepted and scheduled for resolution. The

issues raised by the EASRTP inspections varied in their safety significance,

but were comparable to issues raised during the NRC ASRTP inspection. The team

selected the following significant EASRTP inspection findings (RI) to determine

whether they were correctly scheduled for resolution before restart:

(1) Emergency diesel generator fuse coordination was inadequate (RI 35).

(2) Instrument air lines to the turbine bypass and atmospheric dump valves did ,

not allow for thermal expansion (RI 97) I

i

(3) Integrated control system (ICS) calibration setpoints were not adequately l

controlled (RI 36)

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(4) Decay heat removal (DHR) system reactor building sump isolation valves

were located outside the reactor building and had no covering jackets (RI

18).

(5) Nuclear raw water (NRW) system cooling capacity may not be adequate for

all system heat loads (RI 25).

(6) NRW system flow was not properly balanced (RI 32).

(7) The 120-V ac system cabinet was unqualified because unauthorized ventila-

tion modifications had been accomplished (RI 146).

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(8) Design pressure of MFW heaters was less than pump discharge pressure (RI l

108)

(9) Non-nuclear instrumentation system (NNI) calibration setpoints were

incorrect (RI 150).

(10) Nuclear service cooling water (NSCW) system flows to decay heat removal

(DHR) coolers were above the design maximum and could cause tube damage

(RI104).

(11) NSCW system parallel flow analysis was incomplete (RI 107).

(12) NSCW heat removal capabilities may not be adequate for DHR system loads

(RI 128).

(13) No technical bases were stated for response time acceptance values for the

anticipatory reactor trip system (ARTS) surveillance tests (RI 194). l

(14) Carbon steel components were used in the relief valves of a borated water

system (purification and letdown) (RI 190).

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(15) Containment fire protection system capabilities appeared inadequate

(RI 210).

In each case, the licensee had scheduled the resolution of the selected findings

before restart. The inspection team did not review the licensee's corrective

actions taken to resolve the EASRTP inspection findings, as they were still

being implemented. The tracking and closeout of EASRTP findings was being

conducted by the licensee's Engineering Response Team.

3.1.2 Review of EASRTP Methods

The NRC team interviewed EASRTP team members and reviewed document sheets that

identified the areas inspected for each system. The NRC sampled the EASRTP

inspection documentation for the following systems:

(1) emergency diesel generator l

(2) ac electrical

(3) fire protection

(4) nuclear raw water

(5) nuclear service cooling water

(6) main feedwater

(7) decay heat removal

In general, the methods used to conduct the EASRTP inspection were comparable

to that used for the NRC ASRTP inspection. However, the NRC team identified

the following areas in which enhancements could improve the EASRTP process:

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(1) Some aspects of the operating, testing and maintenance programs were not

ready for the EASRTP inspections. Several modifications, procedures, and

work items were incomplete, preventing these functional areas from being l

fully evaluated. The EASRTP inspectors often had only draft procedures or  :

I

preliminary guidance available to assess these functional areas for

several key systems. In these cases, the inspectors provided valuable l

input to the procedural development process, but were not able to complete

the verification. The EASRTP program did not identify the missing proce-

dures for followup review.

1

(2) The reviews conducted during the EASRTP inspection in the f9' lowing areas

appeared to be less than the reviews performed during the NRC ASRTP

inspection: l

(a) Apparently, only 1 of approximately 50 calculations was reviewed

for the NFW system inspection and problems were identified with that

calculation. The NRC team determined that further calculation  !

reviews should have been made to determine if other problems existed.

(b) The reviews of some surveillance procedures consisted only of

ensuring that major components of the system were covered by testing.

There was no line-by-line review of the surveillance procedure to

ensure that testing was adequate.

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(c) Overload protection for the safety-related 4160-V ac pump motors was

checked by comparing the Maintenance Department records with the ,

specified setpoints identified on drawing E1011. No attempt was made j

to verify the adequacy of the specified relay setpoints. The EASRTP

review of electrical protection for MOVs appeared to only check that

the rated motor load would not trip the circuit breaker. The review

did not consider the overall component coordination or question the

protection for MOV HV20003 which apparently had a 20-amp breaker

protecting a 10-amp wire supplying a 1 amp motor.

(d) The scope of the review for the 125 Vdc system did not appear to

check that the auxiliary building battery charger high voltage alann

setpoint was adequate for the rated voltage range of the de equipment

connected to the bus.

It should also be noted that the EASRTP inspection program covered some

areas in more depth than the NRC ASRTP inspection. The team was favorably i

impressed with EASRTP reviews of the procurement area, walkdown of systems

and assessment of system material condition.

(3) The NRC disagreed with the final disposition of one EASRTP issue. The

EASRTP inspection identified an apparent discrepancy between the nuci t r ,

service cooling water (NSCW) System design-basis document and drawing i

E203, sheet 43. The design-basis document stated that an NSCW surge tank

low-low level alarm would trip the NSCW pump, except when a safety

features actuation system (SFAS) signal was present. Drawing E203 showed

that the NSCW surge tank low-low level alarm would trip the pump with an

SFAS signal present. The EASRTP evaluator, team leader and program

manager concluded that this apparent discrepancy was not significant

enough to require assistance from site engineering to resolve the problem

prior to restart.

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The NRC reviewed this same issue further and determined that the NSCW

surgetanklow-lowlevelswitches(LSLL-48401andLSLL-40402)werenot

safety-related and drawing E203 was correct. However, IEEE-279, "Criteria

for Protection Systems of Nuclear Power Generating Stations," requires

that protective system components be safety-related. Failure of either of

these two level switches in the presence of a valid SFAS signal would

prevent the affected pump from completing its intended safety function.

The inspection team concluded that in this one isolated case, the EASRTP

team did not aggressively pursue their initial observation completely.

The resolution of the apparent problem with the non-safety-related level

switches defeating an SFAS signal to the NSCW system pumps will remain

open pending NRC followup. (50-312/87-29-01)

3.1.3 Comparative Inspection Results

The NRC team conducted an independent review of selected aspects of the systems

listed in section 3.1.2 to form a basis for comparison with the EASRTP find-

ings. For most of the systems reviewed, the NRC inspection team did not

identify any new safety issues. However, some specific concerns were raised

involving the emergency diesel generator system that warrant licensee followup:

(1) The fire door between the emergency diesel generator GEB room and the

east-west hallway did not appear to meet the National Fire Protection

Association Code requirements for separating fire areas. The door must

meet these requirements because cables associated with the emergency

diesel generator GEA run through the east-west hallway outside the door.

This issue will remain open pending NRC followup (50-312/87-29-02).

(2) The drain system for emergency diesel generator GEA and GEB rooms may not i

be adequate to remove the water from the sprinkler systems in the rooms. i

Each room had two 2-inch drain lines that all merged to a 3-inch coninon

drain line. The licensee did not have any analyses to (a) show that the

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two 2-inch drain lines would adequately remove the water from the

sprinkler systems .in the rooms, or (b) that the 3-inch common header would

adequately drain the water from the 2-inch lines. The team was concerned

about the capacity of the 3-inch drain line because there were no check

valves in the drain line to prevent reverse flow. Consequently, water

could back up into the other emergency diesel generator rooms. At the

conclusion of the on-site inspection, the licensee was preparing an

analysis to demonstrate the adequacy of the diesel system for the

emergency diesel generator room. This item will remain open pending

review of the licensee's analyses (50-312/87-29-03).

(3) The design of the starting air system may not be adequate for the new l

emergency diesel generators GEA2 and GEB2. On each new emergency diesel I

generator, the starting air system was safety-related from the inlet

check-valves on the accumulators to the engine. The balance of the air

system was designed as non-safety-related because it was not thought to be

required to perform a safety function after the diesel engine starting

sequence. Upon further review, the licensee determined that air was

required continuously to control emergency diesel generator operations.

In SMUD letter JEW 87-358 to the NRC, dated April 1,1987, the licensee

committed to test the capability of the air receivers to maintain the

required pressure for diesel generator control over a prolonged period of

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time. This item will remain open pending NRC review of the test results

(50-312/87-29-04), l

(4) The air compressor outlet safety valves appeared to be set above the

nameplateratingofthecompressors(290psigvs.250psig). The NRC team

determined that because of the design of the system, this setting could

cause the compressors to operate regularly above tneir nameplate rating.

This item will remain open pending NRC reivew of the resolution of the

discrepancy between the compressor rating and emergency diesel generator

air system operating pressure (50-312/87-29-05).

3.2 Engineering Action Plan Assessment

The Engineering Action Plan (EAP) was developed after the ASRTP inspection

(50-312/86-41) to accomplish the following objectives:

(1) Upgrade engineering practices to improve the quality of future design

work.

(2) Independently review the work perfonned in support of the current outage.

(3) Re-establish the system design bases for the as-built configuration of the

plant.

The EAP was still in the process of being implemented during the NRC inspec-

tion. Report D-0050, "Engineering Action Plan for Rancho Seco Nuclear i

Generating Station," Revision 1, was in the final stages of management review l

before issuance. This report outlined the various programs in progress to '

accomplish EAP objectives and provided the latest draft revisions for design 1

change control procedures. The NRC team assessed the EAP improvements by the

following methods:

(1) Conducted a limited review of EAP design change control procedure

revisions.

(2) Observed the engineering processes being carried out, interviewed SMUD

engineers, and reviewed a sample of engineering documents completed since

the EAP had been implemented. l

(3) Reviewed the results of the independent reviews conducted by Babcock &

Wilcox (B&W) and Bechtel inspectors.

The NRC inspection team noted several improvements in the quality of engineer-

ing activities at Rancho Seco since the initial ASRTP inspection in February l

1987. It appeared that implementation of the EAP was improving the quality of I

design change control activities at Rancho Seco.

3.2.1 Review of EAP Design Change Control Procedures

The inspection team performed a cursory review of the draft procedures and

program overview included as part of the EAP. In general, these procedures and

documents provided adequate guidance for conducting the design change control

process . However, the inspection team identified the following concerns during

its review:

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(1) Draft Procedure RSAP-0301, "Configuration Management Program," Revision 0

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Section 5.2.6, stated that work items may be incrementally closed out

before the entire change package has been completed in order to support

real-time operational needs. The team considered that this practice

should only be allowed after a thorough safety analysis by engineering

personnel takes place to ensure the plant is left in a safe condition.

(2) Draft Procedure NEP 4109, "Configuration Control Procedure," Revision 7,

Section 4.2, stated that engineering change notices (ECNs) were not

required for drawing changes if no actual field work was performed. The

team was concerned that there may be drawing changes that involve no

actual field work but that could have an effect on plant safety.

Potentially significant changes to drawing notes or valve positions would

not require an ECN and the appropriate reviews afforded by the ECN

,

process.

(3) Page 23 of the EAP overview document stated that obsolete calculations

would be deleted. The team was concerned that obsolete calculations

should be superseded, not deleted, in order that a record of the design

evolution is maintained.

(4) There did not appear to be a method for closing out an ECN package that

ensured all the various elements of the ECN were included in the package.

None of the procedures reviewed required an inventory of the required

elements of the package to be listed for package closecut.

3.2.2 Review of the EAP Work Products

The inspection team reviewed the analyses conducted to resolve previous NRC

open and unresolved items, sampled calculations performed by the licensee in

support of new design work, and interviewed engineering personnel to assess the

improvements made by the EAP. Overall, the quality of engineering activities

at Rancho Seco was improving, but the team identified some weaknesses in this

area. The team concluded that these weaknesses were isolated instances and

were attributable to the fact that certain aspects of the EAP had not been

completely implemented throughout the site. The following weaknesses were

identified during the team's review:

(1) During the initial ASRTP inspection, the team identified errors in

calculation Z-FWS-M1742, which justified the acceptability of utilizing a

worn stem nut in an M0V. The licensee replaced the worn stem nut and

repeated the calculation. This new calculation has numerous nonconserva-

tive errors which rendered invalid the conclusion that the worn stem nut

was acceptable. The new errors were as follows:

(a) The thrust load used in the calculation was the minimum thrust

required to allow the valve to operate under design conditions. This

value was less than the thrust generated by several normal modes of

operation that were not considered, such as manual operation, higher

than minimum torque switch settings, and initial valve operation when

the torque switch is bypassed. '

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(b) The effective factor used in the calculation to account for the

uneven load distribution inherent in threaded members was less than

the values identified in the machine design texts referenced in the

calculation. In subsequent conversations with the licensee's

representative, an even lower factor was argued to be acceptable

based on a machine design text. However, the factor cited was for

shear stress whereas the parameter being considered was tensile

stress.

(c) The licensee stated that localized yielding was acceptable, based on

a machine design text. The team agreed that localized yielding was

acceptable for some stati:: applications of screw threads, but it is

not acce) table for a power screw where yielding would cause mismatch

of the t1 reads for subsequent operations.

(d) No consideration was given to the increase in friction and load to  !

produce given thrust as a result of the change in thread angle with

wea r. The team considered that this would likely be the predominant

factor in determining the acceptability of the worn condition.

(e) The originator of the calculation deemed a vigorous analysis unneces-

sary because the operation of the valve had been observed. The team I

disagreed with that conclusion because the observation had not taken l

place at the valve's design pressure drop loading.

Since the stem nut had been replaced, there did not appear to be any

safety problems with the existing system configuration. The team was l

concerned, however, with the programatic implications that this j

calculation may indicate. The calculation appeared to be performed and

'

verified by engineering personnel who were unfamiliar with MOV operation.

Additionally, this calculation was still in the engineering files and

could be used as a model for future calculations of a similar nature. In i

response to the team's concerns, the licensee removed the calculation from I

the files and initiated corrective actions to ensure that similar

calculations will be performed properly in the future. The licensee

maintained that this calculation was an isolated example and not

indicative of the overall engineering program quality. On the basis of

the other engineering activities reviewed, the inspection team agreed with

the licensee's position.

(2) During the initial ASRTP review, deficiencies were identified with the de

short-circuit calculation for the auxiliary building batteries that could

result in the short-circuit current at de panel SOB being above the  ;

manufacturer's interrupting rating for the enclosed circuit breakers

(UnresovedItem 50-312/86-41-16). The licensee replaced the panel feeder

cable with a smaller size conductor to limit the short-circuit current at

the panel. The short-circuit calculation was revised to incorporate the

higher resistance of the smaller cable and also to address other concerns

originally raised by the team. After reviewing this revised calculation,

the team identified mathematical errors that reduced the margin of rated

interrupting current over calculated short circuit by approximately 500

amperes (25 percent of the calculated margin). The calculation cover

sheet indicated that checker had verified the mathematics of the calcula-

tion. The licensee corrected the errors in the calculation and noted that

the breakers were still appropriately sized.

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(3) The NRC inspection team identified the following instances in which there

did not appear to be effective comunications between engineering

personnel and other organizations within the station:

(a) As part of the licensee's response to the original ASRTP concern with

the backup air supply for the EFIC modifications (0 pen Item

50-312/86-41-06), the operating procedure was modified to include

recording the bottle pressures in daily logs. The minimum allowable

pressures specified in the log sheets were at the 2- and 3-hour alarm

points for the bottles which appeared to defeat the purpose of

performing the daily checks. These values were well below the

pressure that had been recormended by engineering personnel in the i

design-basis report for the backup air bottle modifications and did

not appear to have any technical bases. In response to the NRC

concerns, the licensee is revising the operating logs.

(b) A calculation perfomed in response to NRC open item 50-312/86-41-08 l

'

to determine the qualified life of the 0-rings for the excess flow

check valves in the OSTG level instrument lines installed as part of

the EFIC modifications detemined that the expected life of the

components was 10 years. However, this calculated life for the ,

0-rings had not been communicated to the Maintenance Department

for incorporation into its replacement procedures. In response to

the NRC team's concerns, the licensee is changing the preventive

maintenance procedures.

(c) Operations Department personnel were unaware that compressed air was

required for continued operation of the new emergency diesel genera-

'

tor units (GEA2 and GEB2). As a result, the emergency diesel genera-

tor casualty procedures did not include provisions for providing

backup air in the event of loss of the installed non-safety-related

air compressors. This information had not been comunicated from the

Engineering Department to the Operations Department.

The licensee had previously identified similar concerns regarding the

comunication of infomation from the Engineering Department to other

groups at the plant. A group was being established within the Operations

Department to facilitate the implementation of design infomation into

pertinent operating and test guidance.

3.2.3 EAP Independent System Reviews

In addition to the EASRTP inspections and as part of the EAP, the licensee

conducted independent safety reviews of the design, testing, and operation of

the System Review and Test Program (SRTP) selected systems by the following

three programs:

(1) DesignCalculationReview(Bechtel) l

l

(2) Rancho Seco System Configuration and Restart Test Program Evaluation

(Babcock & Wilcox) l

(3) QA Vertical Audit (Stone and Webster)

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These programs duplicated some of the efforts made by the EASRTP teams, but

overall identified a number of significant issues. The NRC inspection team

reviewed the results of these inspections, but not the methods or qualifi-

cation of personnel conducting the H! views. i

(1) The Bechtel Design Calculation Review Program )

I

Those calculations performed in support of the modifications scheduled to

be made during the current outage were evaluated. Of 173 observations

made hy Bechtel teams during their review, 62 required correction before

restart. The licensee determined that the remaining 111 observations did

not require imediate corrective action, but should be considered for

areas of improvement in future calculation revisions. At the time of the

NRC inspection, the licensee had responded to Bechtel on 98 of 173 obser-

vations, outlining its intended corrective action and offering its

proposed schedule. The proposed completion dates appeared to support the

restart date in January 1988, but the team did not review the adequacy of

the planned corrective actions or their implementation.

i

(2) The Rancho Seco System Configuration and Restart Test Program l

1

Thirteen systems designed by B&W that were part of the 33 SRTp selected

systems were evaluated. The NRC inspection team reviewed 66 preliminary i

findings from the evaluation report and found many of these to be

significant issues that should be resolved before restart. Several of the

issues had been previously identified by the EASRTP inspection and actions i

were being implemented to resolve the design problems. The final l

evaluation report was scheduled for issuance from B&W in October 1987 and

final corrective action will be developed on the basis of those findings.

(3) The QA Vertical Audit 4

The system modifications for the EFIC system installation and the electri-

cal portion of the new emergency diesel generator installation were

evaluated. The vertical audit involved a thorough review of the proposed

system design, testing, procurement, and training for the two modifica-

tions. The audit identified several significant issues that must be

resolved before restart. These issues concerned the adequacy of the

design analyses for proposed functional testing of components, operating l

procedures, and quality of components installed in the system. At the

time of the inspection, the licensee's audited organizations had responded

to some of the audit findings. The NRC inspection did not review the

adequacy of these responses or their scheduling with respect to completion

before restart.

The NRC inspection team concluded that these independent reviews of the

selected systems identified issues that were safety significant and contributed

to the overall safe operation of the plant. The team was concerned that the

nature and depth of these findings required significant expenditure of

resources to resolve the issue before restart because many of the problems were

being identified very late in the outage schedule.

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3.3 Procurement and Control of Fasteners

In February 1987, as part of an industry survey, the NRC obtained six fasteners i

from Rancho Seco maintenance personnel that were supposedly safety-related,

quality class I materials. These fasteners were then sent to Idaho National

Engineering Laboratory (INEL) for mechanical, chemical, and microstructural

analyses. As described in an NRC letter (D. M. Crutchfield to G. C. Andognini,

August 26,1987), two of the six fasteners failed to meet the ASTM A-193 Grade

87 standards that were stamped on the bolts. Mechanical testing revealed an

ultimate tensile strength of 54.65 ksi instead of the required 105 ksi;

chemical testing indicated out-of-specification readings for carbon, chromium,

manganese, and molybdenum; and microstructural analysis revealed that the bolts

had not received the proper heat treatment to yield a quench and tempered

martensitic structure as required by ASTM A-193 Grade B7 material. This result

raised concerns that counterfeit fasteners may have been installed in the

plant. The inspection team initially concentrated on the licensee's investiga-

tion into the specific fasteners that had the potential to be counterfeit, but

later expanded the scope of the inspection to include the licensee's program

for procurement and control of fasteners. This effort included further

sampling of fasteners for testing, evaluation of purchasing and control

practices for safety-related fasteners and review of procurement issues raised

during the EASRTP and QA Vertical Audit Programs.

3.3.1 Itvestigation of Counterfeit Fasteners

The two fasteners that failed INEL testing were marked "CB B7," indicating that

they were ASTM A-193 Grade B7 bolts supposedly manufactured by Clark Brothers

Bolt Co. The licensee documented the results of its investigation by SMUD

Office Memorandum MIC 87-186, dated October 3, 1987, entitled "Clark Brothers

t Bolts, ASTM A-193 Grade B7." It appeared that SMUD had purchased 64 "CB B7"

bolts from Westinghouse as non-safety-related, quality class II materials in

1977 and 1982; these were accounted for under stock code 26736. Additionally,

the licensee identified 24 "CB B7" bolts in the warehouse stored under stock

code 25436. The licensee's investigation accounted for all 24 bolts identified

by stock code 25436 and 40 of the 64 bolts identified by stock code 26736. The

licensee committed to locate and replace all the bolts with the "CB B7" head

]

markings.

, After completion of the on-site inspection, the NRC contacted Clark Brothers

Bolt Co. regarding the defective bolts. A preliminary investigation revealed

that Clark Brothers Bolt Co. does not have the capabilities to make the type of

bolts found to be defective at Rancho Seco. Since the bolts were actually

quality class II, the non-safety-related materials were not readily traceable

through Westinghouse.

The inspection team was concerned that the licensee provided quality class II

fasteners to an NRC inspector when quality class I fasteners were requested.

Additionally, the inspection team obtained the following fasteners from the

licensee for further INEL testing:

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NRC Material Meterial BIN Quality

ID No. Description h Stock Code Location Class

RSI ASME SA 193/B7 Bolt 042223 A39LO9 II

RS2 ASME SA 193/B7 Bolt 042223 A39LO9 11

RS3 ASME SA 193/B7 Bolt 042222 A39LO9 II

RS4 ASME SA 193/B7 Bolt 042222 A39LO9 II

RS5 ASTM A449 Bolt 026736 A41F21 II

RS6 ASTM A449 Bolt 026736 A41F21 II

RS7 SAE J429/gr 8 Nut 110340 G02806 I

RS8 SAE J429/gr 8 Bolt 009025 A39J05 II

RS9 SAE J429/gr 8 Bolt 009025 A39J05 II

RS10 SAE J429/gr 8 Bolt 009025 A39J05 II

RS11 304 SS Bolt 008932 A39G09 II

RS12 ASTM F593-85/304 Bolt 103118 A39F08 I

RS13 ASTM F593-84/304 Bolt 103103 A39C07 I

RS14 ASTM F593-84/304 Bolt 103103 A39007 I

RS15 ASTM F593-84/304 Bolt 103103 A39007 I

RS16 304SS Bolt 008932 A39G09 II

RS17 ASTM A193/B7 Bolt Quarantined II

RS18 ASTM A193/B7 Bolt Quarantined II

RS19 304 SS Bolt 008928 A39G09 II

RS20 ASTM F593-85/304 Bolt 103118 A39G09 I

These materials will be tested by an NRC laboratory and the results issued by

separate correspondence.

!

3.3.2 Warehouse Material Controls

The inspection team reviewed the licensee's controls of fasteners that were in

the warehouse while obtaining additional fasteners for testing. The identifi-

cation and segregation of safety-related quality class I fasteners and non-

safety-related quality clans II fasteners appeared to be inadequate. The team i

found the following instarces in which adequate controls did not appear to be

implemented:

(1) Quality class I cap screws, identified in Section 3.3.1 as RS-12 and

RS-20, were found together in the same bin under a single quality accept-

ance (green) tag, but appeared to have been manufactured by different

companies. Both appeared to be stainless steel, but one was polished and

stamped "SS 304" on the head, the other was pitted and not stamped.

(2) Quality class II cap screws, identified in Section 3.3.1 as RS-11 and 1

RS-16, were in the same drawer as quality class I cap screws RS-12 and )

RS-20, but were in separate bins. The two bins were separated by a thin  !

setal divider in the drawer. When the inspectors opened the drawer, the l

loose quality acceptance tag was found in the wrong bin. The cap screws '

in both bins were approximately the same size making it possible to issue

one cap screw in place of the other.

(3) Quality class II cap screws, identified in Section 3.3.1 as RS-5 and RS-6,

were actually ASTM 449 material that was stored in a bin marked for ASTM

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A-193/B7 material. This arrangement presented a problem because ASTM

A-193/87 material has a nominal yield strength of 105 ksi and ultimate

tensile strength of 125 ksi but ASTM 449 yield strength is 92 ksi and an I

ultimate tensile strength in 120 ksi. It appeared that bolts of lower i

!

strength could have been issued from the warehouse for non-safety-related

use. .

l

(4) Quality class II cap screws, identified in Section 3.3.1 as RS-8, RS-9,

and RS-10, were SAE J429 Grade 8 material and were mixed in a drawer with

similarly sized cap screws of ASTM A-325 material. The ASTM A-325 cap l

screws had a tensile strength of 105 ksi and a yield strength of 81 ksi; I

the SAE J429 Grade 8 cap screws, however, had a tensile strength of 105 l

ksi and a yield strength of 130 ksi. The team was concerned that since

there were no markings or tags associated with either of the cap screws,

the wrong material could be issued in error and cap screws of insufficient

strength could be installed for non-safety-related application.

Although the licensee's program for controlling fasteners was not reviewed

thoroughly, the team was concerned that past practices may have resulted in

incorrect fasteners and other materials being installed in the plant for

applications inconsistent with their design.

3.3.3 Purchase Order Review  !

l

The NRC inspection team reviewed the licensee's practices for purchasing 1

'

safety-related fasteners. It appeared that before 1986, the licensee did not

have a written program for the dedication of comercial grade items for  !

safety-related applications or for purchasing quality class I materials or

services. Procedures were issued in 1986 after the licensee discovered this

problem. The inspection team did not review the procedures in sufficient

detail to assess their adequacy. The inspection team sampled the following

safety-related purchase orders:

No. Date Vendor Material I

46520 7/15/83 Cardinal Industrial ASTM A-193/B7 bolts .

Products Corporation ASTM A-194/2H nuts

l

Cardinal Industrial

'

44767 5/12/83 ASME SA 325 nuts and bolts

Products Corporation

16899 3/06/81 Babcock & Wilcox ASME SA 320/C43 studs

ASME SA 194/Gr 4 or 7 nuts

These purchase orders did not impose material traceability requirements,

require work to be perfomed under an approved QA program, or prohibit the

repair or rework as required by 10 CFR 21 and ASME Codes. The team did not l

review whether these purchase order deficiencies resulted in substandard j

material being provided to the licensee.

1

3.3.4 Licensee-Identified Procurement Problems

Several procurement and material control issues also appeared to be identified ,

by the licensee in its EASRTP and QA Vertical Audit inspections. The EASRTP l

inspection identified the following issues which may indicate problems with the

procurement or control of material at Rancho Seco:

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Ref. No. Description j

RI 029, 110 Inadequate comercial grade procurement and

~

dedication methods

RI 068 Carbon steel bolting in systems containing boric

acid

RI 058 Unqualified lube oil coolers in NRW system

Additionally, the GA Vertical Audit Program identified the following concerns  !

during review of the EFIC and emergency diesel generator electrical l

modifications: l

Audit Finding No. Description 1

l

87-20-13 Missing mild environment specification on

purchase orders for NSEB equipment l

l

87-20-17 Inadequate traceability for purchase order for i

target rock solenoid valves for EFIC

87-20-18 Heat number control log deficiencies allowing

for loss of traceability )

87-20-19 QA did not approve all safety-related purchase

orders

87-20-20 Receipt inspection deficiencies

87-20-21 Inadequate QA storage of procurement documents

87-20-22 Missing vendor audit

87-20-23 Inadequate followup for vendor audit findings

The NRC inspection team concluded that there was cause for concern about the

quality of fasteners being used at Rancho Seco Nuclear Generating Station. The

team determined that further review of the procurement area by the licensee and

NRC was necessary based on the EASRTP and QA Vertical Audit findings, problems

identified by the NRC team with the purchasing and warehouse control of safety-

related fasteners, and results of the fastener testing conducted in February

1987. The licensee had already initiated a review of their warehouse materials

to ensure that all quality class I materials were properly documented and

controlled. This item will remain open pending further review and inspection

by the NRC (50-312/87-29-06).

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4 MANAGEMENT EXIT MEETING

An exit meeting was conducted at the conclusion of the on-site inspection on

October 9, 1987. The licensee's representatives at the exit meeting are

identified in Appendix A. Mr. Dennis F. Kirsch Director, Division of Reactor

Safety and Projects, Region V, and Mr. Charles J. Haughney . Chief. Special

Inspection Branch, NRR, represented NRC management at this meeting. The scope

!

of the inspection was discussed and the licensee was informed that the

inspection would continue with further in-office data review and analysis by

tean members. Team members presented their observations for each area

inspected and responded to questions from licensee representatives. The

licensee was informed that some of the observations could become potential

enforcement findings.

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APPENDIX A l

Personnel Contacted

C. Andognini CEO, Nuclear l

J. Anderson Procurement

P. Anderson Electrical Maintenance Engineer

• J. Atwell Licensing

S. Bagga Design Engineer

S. Bajaj I&C Engineer

J. Baldwin EASRTP Evaluator

• M. Basu Lead Electrical Engineer

J. Bergstrom Operations

J. Bingham Civil-Design Engineer

N. Brown Buyer

C. Buchanan Document Control

S. Cannichael EASRTP Engineer

J. Chnastyk Operations

• D. Compton Operations Engineer

J. Cola Preventive Maintenance Manager

• G. Coward AGM, Technical and Administrative Services

• B. Croley Director, Technical Services l

• G. Cranston Manager, Nuclear Engineering

• S. Crunk Technical Assistant, AGM Technical and

Administrative Services

J. Delezinski Nuclear Licensing

Q. Diwan EASRTP Reviewer

J. Dowson QC Coordinator

• D. Falconer Licensing Engineer

• S. Farkas Licensing Engineer

K. Farris Mechanical PM Engineer

• T. Fetterman Supervisor, Electrical Engineering

• J. Firlit AGM Nuclear Power Production

E. Frobom System Engineer

J. Gibson EASRTP Evaluator

R. Gupta Electrical Engineer

R. Gwynn Operation

R. Gynn Operations Engineer

M. Hardin I&C PM Supervisor

J. Hayes Electrical Engineer

T. Himber Assistant Shift Supervisor

J. Hoffman Electrical Maintenance PM Supervisor

R. Holler Procurefrent

• D. Humanansky EASRTP Program Manager

S. Jacobs Electrical Engineer

L. Jau EASRTP Evaluator

P. Johnson Plant Utilities Principal I&C Engineer

J. Kearns Document Control

• J. Kelly Engineering Response Team

• B. Kemper Manager, Operations Department

• D. Keuter Director, Nuclear Operations, and Maintenance

T. Kahn Supervisor, Mechanical Systems

D. Koontz EASRTP Evaluator

• Attended Exit Meeting on October 9, 1987

A-1

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APPENDIX A - (Continued) l

I

Personnel Contacted

• B. Kumar Supervisor-Environnental Qualification 4

F. Lopez Procurement )

R. Lucaro Maintenance i

R. Mayhugh MOV Project Engineer l

J. McColligan Director, Plant Support

D. McIntyre Nuclear Training Department

L. McGreyth Product Quality Engineer

S. Miller Lead Electrical I.ngineer

D. Morena Lead F.lectrical ingineer

D. Morgan Work Planning

J. Myer Procurement

• K. Nyers Manager, Nuclear Licensing

R. Nagel Nuclear Training Department

• M. Nakao Licensing (Bechtel)

T. Parker Quality Assurance Auditor

• M. Price Supervisor, Mechanical Maintenance

• T. Redican Material Management

D. Rice Systems Engineer (AFW)

M. Rojas IDADS

D. Satpathy Supervisor, Mechanical EngineerinD

P. Schwartz Nuclear Training Department

T. Shaw Supervisor-Civil / Structural

• F. Sheenan Lead Electrical Engineer

• J. Shetler Director, SRTP

• T. Shewski Quality Engineer ,

• T. Telford Engineering Response Team

D. Tipton Assistant Operations Superintendent

• R. Tomchuck Procurement

• B. Thomas Public Information

T. Tucker Shift Operations Superintendent

• P. Turner Manager, Nuclear Training

J. Vinquist Director, QA Department i

W. Weaver Procurerrent l

J. Wheeler Senior Electrical Maintenance l

D. Yount ISC/ Electrical Superintendent

- P. Zelmer EASRTP Evaluator ,

J. Zott Group Supervisor-Fire Protection l

R. Zucker Supervisor-0perations and Maintenance

W. Adachi System Design Engineer

J. Brunner System Design Engineer

• R. Nossardi B0P Group Leader

R. Patel I&C Engineer

M. Schlager MOV Engineer Project Manager  ;

K. Srini-Vasiah Design Engineer

• M. Akins EA'iRTP Team Leader

J. Arevalo EA1RTP Reviewer

R. VanEschen EASRTP Reviewer

• Attended Exit Meeting on October 9, 1987

A-2

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APPENDIX A - (Continued)

Personnel Contacted

D. Alemayehu Fire Protection Engineer

H. Grover Electrical Engineer

F. Kayas Electrical Engineer

A. Morales Engineer's Assistant ,

R. Wise Design Engineer-EQ

• F. Stock EASRTP Team Leader

R. Dobson Lead I&C Maintenance Engineer

B. Beebe NSS Principal I&C Engineering

D. Fedol System Design Engineer

• M. Lawrence Engineering Response Team

• E. Murphy Engineering Response Team

T. Marshal EASRTP Team Leader

• X. Prince EASRTP Team Leader

• Attended Exit Meeting on October 9, 1987

A-3

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APPENDIX B

Documents Reviewed

Calculations

Z-VBS-E0523 Rev. 3 07/21/87 120 Vac Load Study

Z-DCS-E0544 Rev. 3 08/14/87 (Aux Bldg.)

7-DCS-E0600 Rev. 3 07/24/87 DC Voltage

Battery Drop (Aux, Bldg.)

Sizing

Z-DCS-E0612 Rev. 3 08/31/87 DC Short Circuit (Aux. Bldg.)

Z-DCS-E0612 Rev. 4 09/30/87 DC Short Circuit (Aux. Bldg.)

Z-DCS-E0636 Rev. 2 09/03/87

Z-DCS-E0732 Rev. 0 07/31/87

Battery

DC ShortSizing

Circuit (NSEB)

(NSEB )

Z-DCS-E0678 Rev. 0 01/04/87 DC Short Circuit (NSEB)

Z-DCS-E0683 Rev. 0 08/07/87 DC Voltage Drop (NSEB)

Z-FWS-E0718 Rev. 0 04/29/87 DC M0V Terminal Volts

Z-EDS-E0744 Rev. 0 09/28/87 Ampacity Derating for H4BAC Cable

Z-VBS-10166 Rev. 0 06/04/87 120 Vac Bus Loading

Z-FPP-E0736 Rev. 0 (Draft) Breaker Coordination

Z-HVS-M2160 Rev. 0 05/22/87 Aux. Bldg. Battery Room Temperatures

Z-HVS-M2129 Rev. O NSEB Battery Room Temperatures

Z-EDS-C0929 Rev. 0 08/07/87 Relay Magnet Mounting Evaluation

Z-FWS-10167 Rev. 0 06/05/87 Aase Calculation for Orifice Plates

Z-ZZZ-C0863 Rev 0 01/21/87 fiM smic Analysis of I&C Valves

Z-EQP-E0689 Rev. 0 01/19/87 Evironmental Analysis of Excess Flow Valves

Z-FWS-IO150 Rev. 1 07/28/87 Required AFW Flow Capability (B&W)

Z-FWS-M2237 Rev. 0 07/14/87 AFW System Head Requirements

Z-ZZZ-C0884 Rev. '2 05/04/87 Condensate Storage Tank Seismic Calculation

Z-MCM-M2212 Rev. 0 09/26/87 CST Overpressure and Vacuum Protection

Z-ZZZ-C0886 Rev. 0 03/03/87 Seismic Calculation for Field Tanks

Z-FWS-M1742 Rev. 1 09/27/87 MOV Stem Nut Strength

Z-DF0-M0119 Rev. 0 03/22/72 Diesel Generator Fuel Tank Sizing l

Z-NRW-M0245 Rev. 0 02/01/74 Spray Pond Level

Engineering Reports

ERPT-E0224 09/29/87 Overcurrent Trip Indication on Westinghouse MCCs

ERPT-E0222 09/20/87 Standby Battery Charger Cable

Studies

B&W Task 847 09/87 Rereview of Instrumentation and Control

for IE Bulletin 79-27

Impell 01-0790-1619 09/09/87 4160 Yac and 480 Vac Switchgear

Impell 01-0790-1620 09/09/87 480 Vac Motor Control Centers

Impell 01-0790-1621 09/09/87 120 Vac Distribution Panels

Impell 01-0790-1622 09/09/87 Alarm and Annunciation Summary Report

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APPENDIX B - (Continued) l

Documents Reviewed l

EASRTP l

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Nuclear Services Cooling Water System  ;

Decay Heat Removal  !

Nuclear Raw Water System )

Emergency Die',el Generator System l

Fire Protection System l

Main Stear System 1

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Purific& ion and Letdown System

React'.,r Building Ventilation and Hydrogen Recombiner

Au,iliary Steam System

125 Vdc System

Integrated Control System l

Control Room / Technical Support Center HVAC System

120 Vac Electrical System

Main Feedwater System ,

Non Nuclear Instrumentation System l

Seal Injection and Makeup System Radiation Monitoring System l

Reactor Protection System

Procedures )

AP.4A Rev. 5 Safe Clearance Procedure Danger Tags 11/14/86

AP.23.00-23.14 Original Conduct of Operations 03/06/87

AP.26 Rev. 13 Abnormal Tag Procedure

AP.90 Rev. 1 Work and Test Authorization Program 09/21/87

MMP-0021 Rev. O Technical and Quality Determination for 03/24/87

Procurement Documents

MMP-0024 Rev. O Identification and Marking of Items 03/18/87

PEP-0025 Preservation, Storage and Maintenance of 09/14/87

Items in Storage

VEP-0026 Rev. O Four Level Procurement System 03/11/87 l

NEAP-4801 Rev. O MOV Design Process

QAP-Policy Rev. O Procurement Document Control 01/01/86

Section IV

QAP-Policy Rev. O Control of Purchased Material. 01/01/87

Section VII Equipment, and Services

QAP-Policy Rev. O Nonconfonning Materials, Parts, or 01/01/86

Section XV Components

QAP 3 Rev. 2 Quality Assurance Classification 01/01/86

QAP 4 Rev. 3 Procurement Document Control 03/19/87

QAP 5 Rev. 1 Supplier Quality Assurance 01/01/86

QAP 17 Rev 4 Nonconforming Material Control 01/09/87

RSAP-0401 Rev. O Preparation and Processing of 03/06/87

Requisitions

RSAP-0704 Rev. O Stock Reclassification Requirements 06/01/87

Evaluation

RSAP-0706 Rev. O Material Control 09/01/87

RSAP-0810 Rev. O ASME Section XI Repair and Replacement 07/27/87

Program

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APPENDIX B - (Continued)

Documents Reviewed

RSAP-1201 Rev. O Procedure Change Training 04/24/87 I

TDI-3321 Rev. O Training Materials Development

A.31 Rev. 28 Diesel Generator System Operation 04/29/87

Procedure

A.31B TDI Diesel Generator System Operating 05/02/87 l

Procedure 1

A.54 Rev. 9 220 Volt AC Electrical System 11/23/87 l

A.62 Rev. 10 120 Volt AC Vital System

C.10 Rev. 2 Main Feedwater Induced Transients 08/31/87

C.143 Rev. 6 Loss of 480 Volt MCC S2E1 03/05/87

SP.300 Rev. 6 Weekly Nuclear Service Battery Pilot 08/26/87 ,

Cell Test I

SP.301 Rev. O Monthly Nuclear Service Battery Test 08/26/87

STP.1070 AFW Hot Shutdown Test 09/14/87

STP.1086 AFW Pump Flow Test with Condenser at 08/20/87 ,

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Atmospheric Pressure

Drawings

S20E4GW-0710-000 Rev. 5 05/19/83 GM Diesel Air Start System ,

520E4GW-0425-000 Rev 5 04/03/84 GM Diesel Coolant Flow Diagram  !

S20E4GW-0600-000 Rev. 3 08/13/81 GM Diesel Fuel Oil Schematic

S20E4GW-0300-000 Rev. 4 07/29/81 GM Diesel Lub Oil System i

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M-582, Sheet 1 Rev. 24 03/14/79 GM Diesel Oil System

M-583, Sheet 1 Rev. 11 09/25/86 GM Diesel P&I Diagram

M-522, Sheet 1 Rev. 35 08/12/87 Decay Heat Removal P&ID

M-545 Rev. 18 08/20/87 Nuclear Service Cooling Water P&ID

M-544 Rev. 21 08/20/87 Nuc. Service Raw Water System P&ID

M-342, Sheet 1 Rev. 16 08/30/85 Aux. Bldg. Flow Drains

M-342, Sheet 2 Rev. 1 05/20/87 Aux. Bldg. Flow Drains

A-110, Sheet 1 Rev. 19 Aux. Bldg. Grade Level Plan

Conesco 5675-1 Rev. 2 04/08/81 Diesel Oil Storage Tank

M-585, Sheet 1 Rev. 0 10/31/86 TDI DG-Train A

M-585, Sheet 2 Rev. 0 10/31/86 TDI DG-Train A

M-585, Sheet 3 Rev. 0 10/31/86 TDI DG-Train B

M-585, Sheet 4 Rev 0 10/31/80 TDI DG-Train B

M-547 Rev. 0 10/31/86 New Diesel Fuel Oil System

E-1012, Sheets 101, 102, 114, 150, 156, 159, 171, 187, 200, and 268

Westinghouse Electric Corp. Final Assembly Drawing-Feedwater Heater

Material Requests

Bolt. Hex Head,1.00 Dia x 12.00 length 10/20/76

Bolt, Hex Head, 5/8" x 2-3/4", CB Material 02/28/77

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APPENDIXB-(Continued)

Documents Reviewed

Documentation Concerning CB-B7 Bolt

Partial Receiving Record #55101 09/19/76

Partial Receiving Record #55117 10/09/76

Partial Receiving Record #54189 08/16/76

Partial Receiving Record #54175 08/05/76

Shipment Notice SF-81155-T3 07/30/76

Partial Receiving Record (illegible) 07/22/76

Shipment Notice SF-81155-T3 07/21/76

Partial Receiving Record #55047 07/12/76

Shipment Notice SF-81155-T3 07/10/76

Partial Receiving Record #54161 07/19/76

Shipment Notice SF-81155-T3 07/17/76

Purchase Orders

RS 35449 Nut, Hex, S/8" and Bolt, Hex Head 08/25/82

5/8" x 2-3/4", CB Material

RS 30269 Studs for Check Valve, Size 3/4" x 4-1/2" 04/21/82

RS 4476 5/8" dia x 3" long Hex Bolt w/ Heavy Hex Nuts 05/12/83

RS 46520 ASTM A193 6R B7 Stud Bolts 5/8" diameter 07/15/83

RS 16899 Studs (1-14"-8UN-2A), Nuts (1-1/4"-8UN-28) 03/06/81

Secondary Manway

RQ-87-08-55800 Nuts All Thread Rod, Cap Screws, and Bolts 06/03/87

RS 74248 Stud 4" x 8 thd x 32" long, and Nut 4" x 8 thd 06/21/87

Work Requests

1. 134329 Number 1 & 2 Low Pressure Turbine

1. 134330 Service Spare 550 MVA XFFR

1. 134331 Service Desuperheater Inlet, Air Operated Valve

Miscellaneous

Special Order 87-19, "Operations Audits"

Training Department Memo PET 87-169

Memorandum JVV 87-055, Bulk Items Exempt from Green

Tagging Requirement (09/30/87)

Memorandum--Revised P0 Number Format. Inventory Account, and Initials

and Codes found in NUCLEUS (12/20/86)

Report No. D-0050 Engineering Action Plan plus Appendices A, B, C, and D

(09/04/87)

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APPENDIXB-(Continued)

Documents Reviewed

Design Casis Report for ECN-R-1188, Rev. 0 (06/03/87)

USAR, Section 9.9, Fire Protection

SMUD Office Memorandum - Subject: Clark Brother Bolts ASTM A193,

Grade B-7 (10/03/87)

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