ML20206N631

Insp Rept 50-312/86-21 on 860602-0711.Violation & Deviation Noted:Auxiliary Feedwater Flow Indication in Control Room Not Powered by Class 1E Power Supply,Nor Built to Class 1E Requirements

ML20206N631
Person / Time
Site:	Rancho Seco
Issue date:	08/07/1986
From:	Miller L, Myers C, Perez G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20206N592	List: ML20206N588 ML20206N596 ML20206N607 ML20206N631
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM 50-312-86-21, IEB-85-003, IEB-85-3, IEIN-85-074, IEIN-85-74, IEIN-86-029, IEIN-86-29, IEN-86-29, NUDOCS 8608260330
Download: ML20206N631 (15)

See also: IR 05000312/1986021

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No: 50-312/86-21

Docket No. 50-312

License No. DPR-54

Licensee: Sacramento Municipal Utility District

P. O. Box 15830

Sacramento, California 95813

Facility Name: Rancho Seco Unit 1

Inspection at: Herald, California (Rancho Seco Site)

Inspection conducted: '

Inspectors: _

n r 'sident Inspector

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Date Signed

C. J{/Mpyr , Acti

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G. P.(Pgf , Reside - nspector

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Date Signed

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L. F. Hillbr, Chief, Resctor ProjectsSection II 6 ate Signed

Summary:

Inspection between June 2 and July 11, 1986 (Report 50-312/86-21).

Areas Inspected: This routine inspection by the Resident Inspectors involved

the areas of operational safety verification, maintenance, surveillance,

training, safety system walkdown, Bulletin review, management meetings, and

followup items. During this inspection, Inspection Procedures 30702, 30703,

36700, 41400, 41701, 61725, 61726, 62702, 62703, 71707, 71710, 90712, 92700,

92701, 92702, 92703, 93702, 94702 and 94703 were used.

Resulta: Of the areas inspected, one violation and one deviation were

identified. The violation concerned implementing NUREG 0737, Item II.E.1.2,

safety grade auxiliary feedwater indicatioa in the control room. The

deviation concerned the lack of continuous cleaning of the nuclear service raw

water.

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DETAILS

1. Persons Contacted

Licensee Personnel

J. Ward, Assistant General Manager

G. Coward, Manager, Nuclear Plant

• J. McColligan, Assistant Manager, Nuclear Plant

D. Army, Nuclear Maintenance Manager

• B. Croley, Nuclear Technical Manager

D. Gillispie, Nuclear Engineering Department, Manager

S. Redeker, Nuclear Operations Manager

J. Shetler, Nuclear Scheduling Manager

• T. Tucker, Nuclear Operations Superintendent

M. Price, Nuclear Mechanical Maintenance Superintendent

L. Fossom, I&C Maintenance Superintendent

R. Colombo, Regulatory Compliance Superintendent

J. Field, Nuclear Technical Support Superintendent

S. Crunk Incident Analysis Group Supervisor

J. Jurkovich, Site Resident Engineer

F. Kellie, Radiation Protection Superintendent

L. Schwieger, Quality Assurance (QA) Manager

M. Hieronimos, Assistant to the operations Superintendent

J. Jewett, Site QA Supervisor

H. Canter, QA Operations Surve111ence Supervisor

• C. Stephenson, Regulatory Compliance Engineer

-B. Daniels, Supervisor, Electrical Engineering

J. Irwin, Supervisor, I&C Maintenance

• C. Linkhart, Electrical Maintenance Superintendent

• Q. Coleman, QA Acting Site Supervisor

Other licensee employees contacted included technicians, operators,

mechanics, security and office personnel.

• Attended the Exit Meeting on-July 11, 1986

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2. Operational Safety Verification

The inspectors reviewed co'ntrol room operations which' included access

control, staffing, observation of decay heat removal system alignment,

and review of control room logs. Discussions with the Shift Supervisors

and operators indicated understanding by these personnel of the reasons

for annunciator indications, abnormal plant conditions and maintenance

work in progress. The inspectors also, verified, by observation of valve

and switch position indications, that emergency systems were properly

aligned for the cold shutdown condition of the facility.

Tours of the auxiliary building. turbine building, and reactor building,

including exterior areas, were made to assess equipment conditions and

plant conditions. Also the tours were made to assess the effectiveness

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of radiological controls and adherence to regulatory requirements. The

inspectors also observed plant housekeeping and cleanliness looked for

potential fire and safety hazards, and observed security and safeguards

practices.

The inspector noted a series of recent events indicating a lack.of

attention to detail in the performances of routine personnel activities:

A. Two pipe snubbers were mistakenly removed for testing from.an

operating train of the decay heat removal system (described in

Paragraph 4B).

B. The valve operator motor for the pressurizer EMOV was burned out

when mistakenly operated for a surveillance prior to completion of

maintenance testing and adjustment of the operator limit switches.

This item is unresolved pending further review (86-21-04).

C. A train of the decay heat removal system was put into service

without initiating operation of the nuclear service raw water system

as required by plant procedures. This item is unresolved pendin;

further review (86-21-08).

D. A valve line-up error when sampling the Once Through Steam Generator

(OTSG) secondary chemistry resulted in inaccurate analysis and _

excessive hydrazine additions.

. At the exit interview, the inspector emphasized the need for increased

attention to detail in all areas of plant performance in light of the

increase in maintenance activity that the plant is conductive restart

activities. Licensee representatives acknowledged that lack of attention

to detail was a concern of SMUD, and appropriate corrective action for

each of these occurrences would be taken. ,

3. Maintenance

Several maintenance activities for the systems and components listed

below were observed and reviewed to ascertain that they were conducted in

accordance with approved procedures, regulatory guides, industry codes or

standards, and the Technical Specifications.

The following items were considered during this review: The limiting

conditions for operation were met while components or systems were

removcd from service; approvals were obtained prior to initiating the

work; activities were accomplished using approved procedures and were

inspected as applicable; functional testing or calibration was performed

prior to returning components or systems to service; activities were

accomplished by qualified personnel; radiological controls were

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implemented; and fire prevention controls were implemented.

A. Cabinet Work

The inspector observed work in the safety cabinets at various timeo

during thin report period. This work involved lug replacement, wire

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wrap, and soldering. The cabinets involved were the reactor-

protection and the novi-maclear instrumentation cabinets.

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Y",, No violations or deviations were identified.

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q(/ B. MOV Inspection'

The inspector observed portions of maintenance activities to

determine the as-found condition of.30 Limitorque operators in

response to IE Bulletin 85-03 (described in Section 9 of this

report). Partial results of the licensee's inspection indicate

several areas of deficiencies in the operators.

The licensee.will evaluate the data collected during these

inspections to assess the significance of the deficiencies,'and

whether it is appropriate to expand the' scope of the Limitorque

operator inspection'and refurbishment program.

The inspector noted that discrepancies found during the licensee's

maintenance inapection were not documented on.a Nonconformance

Report;(NCR), but rather were documented on the serialized work

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request.- This is a repeat of a'similar concern expressed by the

inspector in review of maintenance activities involved in replacing

cracked reactor coolant pump bearing cap screws during inservice

inspection.(Unresolved item 85-25-02: OPEN). The inspector

questioned whether this was an appropriate means of' control of the

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discrepant conditions to insure adequate review of the findings and

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proper disposition. The. inspector determined that the licensee's

procedure AP.3 " Work Request" allows annotation of the work request

for rework or replacement activities which are performed on Class 1

systeme or components to restore proper functioning within

specifications. 'This alternative to NCR procersing also requires-

serializing of the work request by QA to provide audit capability.

Cognizant licensee Quality Department-personnel stated that the.

intent of the serialized work request was to allow minor changes to

the scope of the approved work request to deal with unanticipated

maintenance findings which do not require failure analysis in

determining appropriate dispositions. A licensee representative

stated that an Occurrence Description Report (ODR) would be written

addressing all the findings documented on the serialized work

.requer,ts to insure that any generic consequences are appropriately

evaluated.

This item will remain unresolved pending review of the licensee's

ODR and evaluation of the equivalence of the use of serialized work

requests in controlling non-conforming items of maintenance. Of

particular concern is the lack of documentation of the cause and

corrective action for nonconforming condition on serialized work

requests.

As a result of the December 26, 1985 overcooling event, several

preventive maintenance program weaknesses discussed below have been

addressed by the licensee. The inspector monitored the progress of

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changes to the licensee's maintenance program which were being

implemented as part of the licensee's restart program.

The licensee has consolidated their maintenance departments under a

single Nuclear Maintenance Manager and authorized moderately increased

staffing within the department to establish dedicated resources for

preventive maintenance. To date, actual staffing has increased slightly.

The licensee is currently pursuing INPO accreditation of their

maintenance training program. The inspector observed increased attention

and emphasis in the area of maintenance training demonstrated in

preparation for motor operated valve operator inspections as part of the

licensee's program in response to Information Bulletin 85-03.

Ongoing inspections of manual valves to insure operability under

off-normal events has been expanded based on valve conditions revealed by

these inspections and will provide the basis for defining selection

criteria for expanding the scope of preventive maintenance on plant

valves.

The inspector found that the changes underway were.significant

improvements in the licensee's preventive maintenance program with the

potential for achieving a demonstrated improvement in system and

equipment reliability when fully implemented.

No violations or deviations were identified.

4. Monthly Surveillance

Technical Specification (TS) required surveillance tests were observed

and reviewed to ascertain that they were conducted in accordance with

these requirements.

The following items were considered during this review: TestinE was in

accordance with adequate procedures; test instrumentation was calibrated;

limiting conditions for operation were met; removal and restoration of

the affected components were accomplished; test results conformed with TS

and procedure requirements and were reviewed by personnel other than the

individual directing the test; the reactor operator, technician or

engineer performing the test recorded the data and the data were in

agreement with observations made by the inspector, and that any

deficiencies identified during the testing were properly reviewed and

resolved by appropriate management personnel.

A. 125V DC Station Batteries

The inspector reviewed the 'A' battery discharge test conducted on

June 23, 1986. No discrepancies were noted at the time. However,

while the system was being aligned to its normal configuration the

following occurred: The 'A' battery charger (H4BA) tripped and

deenergized the SOA bus. This was unexpected because the standby

battery charger (H4BAC) was powering the SOA bus prior to the

occurrence; six panel annunciator lights burned'out on the 'A' and

'B' diesel generator panels; the 'A' battery charger subsequently

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was reset and tripped four more timas, but in these cases the

standby charger still carried the bus.

The licensee in response to this occurrence declared both the 'A'

and 'B' diesel generatore inoperable because of the unexplained

light failures. The normal surveillance tests for the diesel

generators were performed, the diesel generators passed, and were

declared operable. No cause for the light failures or the failures

of the standby battery charger to carry the 'A' bus was determined.

At the end of this report period the licensee had developed a draft

Troubleshooting Action Plan. The inspector discussed with the

licensee at the report exit that timely troubleshooting of this

event was considered important. Therefore, this item would,be

considered an open item until the troubleshooting action plan has

been developed, implemented and corrective actions accomplished for

this event. This is an open item (86-21-01)

B. Snubber Testing

The inspectors observedIpertions of the snubber testing performed

under SP 201.10B " Safety System Hydraulic' Snubber Functional

Testing" and MT.020 " Snubber Functional Testing". On June 26, 1986

the inspectors observed the failure of the surveillance test of'

snubber #129 (1-SW-23822-7B). 'During'the extension test the snubber

failed to lock-up., The next day, June 27 the inspector was

informed that snubber #129 had been tested again and passed its

surveillance test. The licensee declared the snubber operable. The

inspector did not find,any written determination to justify a second

test or to accept the resuite of the passed second test over the

failed first test. In addition, Quality Assurance Procedure QAP 26

" Test Control" provides the requirements for Acceptance and

Surveillance Tests. The procedure required an NCR to be generated

.when an acceptance test result is not in conformance to the

acceptance criteria, but it did not provide a similiar requirement

for a failed surveillance test. QAP 26 appears to be incomplete in

dealing with surveillance tests.

After the inspector discussed his concerns with the licensee, the

licensee issued a NCR on July 8, 1986 which documented the events of

June 26 and 27. The NCR had not been dispositioned by the end of

this report perfed. The inspector considers the occurrence

discussed tbove as an Unresolved Item, 86-21-02.

In further reviews of the snubber testing program by the inspector

additional concerns had been raised. These concerns were the-

following:

(1) The licensee identified in a different NCR that the technical

specifications required a snubber hydraulic seal replacemnt

program, such that the snubber seals would be replaced before

their expected service life is exceeded. At the end of this

inspection, the service lifetime had not been defined. The

-f quality assurance department had determined that the service

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- life for the seals was five years, and had identified

sixty-four safety-related snubbers with no documentation for

seal replacement within the past five years. The NCR was

awaiting disposition at the end of this report.

(2) The refueling interval functional test requires that ten

percent of the total of each type of snubber in use in the

plant be selected. At the end of this report period, the

inspector had not determined if the initial representative

sample had been composed properly.

(3) A member of the licensee staff, when asked bp the inspector

what the justification was for accepting snubber #129 as

operable after it had failed its first test, replied that there

existed a temperature dependency of the snubber's lock-up

velocity. Therefore, the snubber reportedly failed due to high

temperature conditions during the test. However, this  ;

justification was not documented nor was any explanation given

for the possible effect of temperature on the snubber in

operation.

These items will be reviewed.and inspected in conjunction with the

already discussed unresolved item.

However, during this report period the licensee identified that on

June 6 and June 9, 1986 two snubbers were removed from the operable

'B' decay heat system (DHS) train. Technical Specification 3.12

" Snubbers," requires, in part, during a Cold Shutdown condition

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(which the plant was in during this report period) for snubbera

located on systems required to be operable for nuclear safety, "With

one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore

, the inoperable snubber (s) to operable status and perform an

- engineering evaluation per specification 4.14.c on the supported

component or declare the supported system inoperable and follow the

appropriate specification statement for that system." The licensee

did not replace or restore the snubbers within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period or

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perform the engineering evaluation or declare the 'B' DHS

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inoperable. This is an apparent violation of Technical

Specification 3.12. The licensee did not recognize the conditien

until June 26, 1986, at that time the snubbers were functionally

I tested and reinstalled.

The licensee performed an Incident Critique (#IC 86-04) of the above

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event and recommended specific actions to preclude recurrence of

this type of event. The inspector reviewed'the recommendations and

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The' inspector found that the licensee had identified the problem, it

was reported, via 10 CFR 50.72, the condition was corrected, an

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Incident Critique was performed in a timely manner which included

i measures to prevent recurrence, and this occurrence apparently was

l not a repeat violation. Therefore, the violation is considered

. licensee identified.

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5. Safety System Walkdown

The inspectors verified the operability of the following systems required

in a cold shutdown condition: Decay Heat (DHS), Nuclear Service Raw

Water (NSRW), and Nuclear Service Cooling Water (NSCW).

The inspectors performed system walkdowns and found that the system

line-ups were in accordance with the licensee's operating procedures and

matched the as-built drawings. The inspectors verified that the valves

were in proper position, with power available, and the valves that were

required to be locked were found locked. Control room indication of the

system's equipment was compared to the local indication and was found

adequate.

,The inspectors assessed the equipment conditions and housekeeping of the

DHS and the NSCW system and found it to be adequate. However, the

material condition of the NSRW, including the spray ponds, was found to

be degraded. The licensee's Updated Safety Analysis Report (USAR),

Section 9.4.2.1, " Nuclear Service Raw Water System," states in part:

" Nuclear service raw water is treated with hypochlorite to prevent algae,

slime, and bacterial growth, and with corrosion inhibitor. Each of the

ponds has a separate chemical feeding and filtering system for continuous

cleaning of the water." On various occasions and specifically on

June 23, 24, and 25, 1986 the two spray pond filters were found not

operating and not providing continuous cleaning. This is a deviation to

the USAR commitment for having continuous cleaning of the spray ponds.

Deviation (86-21-03).

The inspector also found organic growth in the corner areas of the spray

ponds, two partially-plugged spray nozzles on the east spray pond,

deficiency tags on a circulating pump which indicated the pump was

out-of-service, and a layer of crust on top of the east filter's sand

bed, indicating the filter had not been in use for a period of time. The

inspector brought these observations to the licensee's management. The

licensee immediately began an investigation of the NSRW system. Their

findings were as follows: Twenty-two open work requests on the NSRW

system including vegetation growth in spray ponds, two partially-plugged

spray nozzles, obstructed lines in one of the hypochlorite systems, an

out-of-service circulating pump, a filter unit in continuous backwash, an

ammeter out-of-service on a NSRW pump, and several minor leaks. The

licensee, however, determined the NSRW systems to be operable. The

inspector discuesed with the licensee the importance of maintaining the

material condition of the NSRW system and other systems required for

plant operation. It is apparent that a large backlog of work requests

existed at the end of.this report-period, and that if the work requests

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are not accomplished in a timely manner it could eventually affect the

operability of safety systems.

The inspector identified one deviation (86-21-03) and various concerns

about the NSRW system's material condition. These items will be reviewed

with the licensee's corrective actions taken on the deviation.

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6. Followup Itsms

Enforcement Items

83-34-01/83-34-02 (CLOSED) - These violations were two examples of

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the licensee's failure to follow procedures. In both cases a

temporary change to a procedure was made which altered the intent of

the procedure. This is prohibited by the licensee's technical

specifications.

The licensee agreed with the above violations, issued a memorandum

for the staff on the' proper use of temporary changes, and revised

the procedure for procedure control to clearly identify the proper

use of temporary changes. ,The inspector reviewed these actions and

found them adequate. In addition,' the inspector had conversations

with operators and engineers about temporary changes and found their

knowledge adequate. .Therefore, the above items are closed. CLOSED

(83-34-01 and 83-34-02).

IE Information Notice IN No. 85-74: Station Battery Problems

(Closed)

The notice described problems that have occurred with lead-acid

station batteries at several nuclear power plants. The inspector

has performed inapections of the licensee's station batteries

(Inspection Reports 85-08 and 86-07) and has documented similar

problems. The problems identified are being followed through the

normal inspection program. The licensee is also in the process of

replacing the station batteries. This will be completed prior to

restart.

The inspector discussed IN 85-74 with a licensee representative. It

appeared that the appropriate personnel at the site received and

reviewed the Notice. The licensee's action to address the

Information Notice appeared satisfactory, therefore, IN 85-74 is

considered CLOSED.

Licensee Event Reports (LER)

LER 85-10 and 85-10, Revision 1, (CLOSED) - This report described

the high point vent leak which occurred on June 23, 1985.

Inspection of this event was documented in Inspection Report 85-19

and a Notice of Violation was issued on September 26, 1985. The

licensee's corrective actions were detailed in the LER and in the

revised LER. The corrective actions appear to be complete and

appropriate to prevent reoccurrence of this event. Due to

inspections performed previously this LER 85-10 and 85-10,

Revision 1 is CLOSED.

7. Licensed and Non-Licensed Operator Training

The inspector continued an inspection of the licensed and non-licensed

operator training programs that had started and was documented in

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Inspection Report 86-07. The inspector reviewed records for a selection

of senior reactor operators and control room operators and verified the

records contained the following: Copies of the most recent annual

written examination and the individual's responses, documentation of

attendance at required lectures, documentation of control manipulations

at the simulator, a copy of recent performance evaluation, and

documentation that required readings had been received and completed.

The inspector also looked at the pass rates for requalification exams

given by the licensee. For the control room operators, the pass rates

were 90% in 1984 and 100% in 1985, and for.the senior reactor operators,

the pass rates were 94% in 1984 and 100% in 1985. During the same period

the pass rates were calculated for initial licensed operator exams are as

follows: 100% in 1984 and 75% in 1985 for the reactor operators and 100%

in 1984 and 85% in 1985 for the senior reactor operators.

The inspector reviewed the training requireme'nts for a selection of

non-licensed operators, and found the records to reflect the training

given to the operators. Discussions were held at various times with the

non-licensed operators to verify that training had been given.

The inspector verified that the licensee received INPO accreditation in

April of 1986 for the senior reactor operator, reactor operator, shift

technical advisor, and non-licensed operator training programs. The

licensee's remaining training programs (maintenance, chemistry and

radiological protection, technical support, and managers) have been

submitted to INPO for accreditation.

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The inspector reviewed a sample of training records and found that the

licensee is meeting the regulatory requirements for qualification of

various staff members.

No violations or deviations were identified.

8. NUREC 0737 Action Item. II.E.1.2 Followup

Item II.E.1.2 required the licensee to install both safety grade

automatic initiation of the Auxiliary Feedwater System (AFW) and safety

grade indication of auxiliary feedwater flow to each steam generator in

the control room. Item II.E.1.2.2 is described in NUREG 0737 as:

" Position

Consistent with satisfying the requirements set forth in General

Design Criterion 13 to provide the capability in the control room to

ascertain the actual performance of the AFWS when it is called to

perform it's intended function, the following requirements shall be

implemented:

1. Safety-grade indication of auxiliary feedwater flow to each

steam generator shall be provided in the control room.

2. The auxiliary feedwater flow instrument channels shall be

powered from the emergency buses consistent with satisfying the

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emergency power diversity requirements of the auxiliary

-feedwater system set-forth in Auxiliary Systems Branch

Technical Position 10-l'of the Standard Review Plan,

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Eection 10.4.9."

In addition Criterion 13 of 10 CFR 50, Appendix A " General Design

~;. Criteria for Nuclear Pcwer Plants" states in part: " Criterion 13 -

Instrumentation and Control. Instrumentation shall be provided to

monitor variables and systems over their anticipated ranges for normal

operation, for anticipated operational. occurrences, and for accident .

conditions as appropriate to assure adequate safety..."

The NRC issued an Order, 7590-01, confirming commitments to implement

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certain post-TMI items. The Order, 7590-01, dated March 14,~ 1983,

stated in part:

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"II.E.1.2 Auxiliary Feedwater (AFW) Initiation and Flow Indication

Safety grade AFW initiation and flow indication is part of an

expanded AFW system upgrade at Rancho Seco. The upgrade will

include modifications (safety grade control) that are beyond those

suggested in NUREG-0737, and accordingly the expanded upgrade will

not be~ completed until the 1984 RO. However, the safety grade

1 initiation and flow indication portioi will be installed during the

! February 1983 RO. Late delivery of equipment contributed to the

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delay.- Existing control grade AFW initiation and flow indication'

provide adequate interim protection."

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The inspector reviewed Engineering Change Notice (ECN) A-3622,

Revision 2, which was the work package for the AFW flow indication

i modification required by item II.E.1.2. The ECN package was signed as

completed on July 8, 1983. In the Design Basis Report (DCN) of

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ECN A-3622, Revision 2, the following was found:

I. " PURPOSE OF DESIGN CHANCE:

, The purpose of this design change is to implement the requirements

! of NUREG 0737 Item II.E.1.2.2 auxiliary feedwater flow

indication....

This is further defined in NUREG 0737, II.E.1.2 Part 2, by stating

that safety grade auxiliary feedwater flow indication is required to

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meet the 10 CFR part 50, Appendix A Criterion 13.

1. Safety-grade indication of auxiliary feedwater flow to each

steam generator shall be provided in the Control Room...

II. DESIGN CRITERIA USED:

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3. Signals to control Room for indication will not be IE in

interim modification during the 1983 outage, therefore, the

indication in the Control Room is not Class IE..."

The inspector reviewed plant drawings (I-53, Sh.6, N25.01-33,

Ravision 10, N28.02-2, Sh.1, Revision 0, I-647 and E-402, Sh.7, and 8.)

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. and interviewed; licensee representatives'from Nuclear Engineering. -

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Nuclear Licensing, Compliance, and Quality Assurance and found :that the ~

indication of AFW flow in the control room was in fact not Class IE. Not

having Class IE power to the APW indication in the control room does not

I provide the availability'of the instrument for anticipated operational

occurrences, and'for ac'cident conditions; for example,. loss of offsite

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power.1 Therefore, it. appears that criterion.13 was not met;.furthermore,

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the TM1 iten II.E.1.2 was also not met. Therefore, it appears that from ,

the time that the modification was completed,-July ~8, 1983, through the

end of this report period, the licensee has been in vio.lation of NRC

Order 7590-01. This is an apparent violation. (86-21-05)

9. IE Bulletin 85-03 Followup

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'The inspector monitored the licensee's test and corrective actions which

were initiated during this inspection period in response to 4 [

IE'Eulletin 85-03 ." Motor-Operated Valve Common Mode Failures During ~

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r _ Plant Transients Due to Improper Switch Settings," dated

November ~15, 1985. In their response dated May 16, 1985, the licensee

committed to develop and implement a two phase program to ensure that

valve operator switches of motor operated valves (MOVs) were specified,

i set and maintained properly.. The first phase reviewed the" design basis

1 :for the operation of motor operated valves in the high pressure injection.

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(HPI) and auxiliary feedwater systems (AFS) to document the maximum ,

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differential pressure-expected on these valves during both normal and' - _

, ' abnormal events. <The second phase of the' licensee's program' involved six

steps!.

e

1) Valve data collection to establish as-found conditions -

4

2) Operator refurbishment ~

3) Analytical determination of the valve stem thrust developed by

the' motor-operator based.on manually. applied' torque

4) Setting of.the ' switches using4the analytical valves '

5) Testing to confirm the valv,e will perform.when required to

mitigate an accident. i -

' -

_ .

6) Revision of procedures to ensure that the switches are correctly

4 set and maintained. -

-

In addition, the licensee's program committed to incorporate the items of

IE Notice 86-29, " Effects of Changing Valve Motor-Operator Switch '

Settings." t ,

a

!' Atthetimeoftheinspectiob,the'licenseehadcompletedthefirstphase

of their program and a training program for Maintenance and Quality ,

i. ' Department personnel who would be involved in subsequent phase two motor

operator work. The licensee identified 30 Limitorque operated MOVs in

the HPI/AFS systems to be investigated under their program and determined

4

-the maximum differential pressure under which each would be required to

, operate in a design basis event.

The inspector reviewed the valve actuator training program conducted ~by.

Power Safety Internationalsfor the licensee personnel and.found the

1 program to be extensive and comprehensive, presenting instruction on

t

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.

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.

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.

12 '

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, maintenance, repair and troubleshooting of Limitorque design valve

-

< -

, operators.

.

, The inspector found that licensee's response to IEB 85-03 had been timely

cnd that they were implementing their program in accordance with.their

'

scheduled commitments.

i

'

During the period of the inspection, the-licensee initiated phase two of

their program to collect as-found valve data as input for their

analytical determination of valve stem thrust forces.

The inspe'etor discussed the. licensee's program with' cognizant licensee

personnel, reviewed the licensee's work procedures controlling the

implementation of their program and observed portions of the

'

investigative activities performed and found them to be consistent with

. the committed program (except as noted below).

4

- During this review the inspector made the following observations:

1. No previous preventive maintenance program had been established

'for control and maintenance of the Limitorque operators. A

licensee procedure (EM.ll7,' Revision 5, " Functional Test of

>

Valve Motor Operators") provided instructions on-setting the

-' operator switches for use during corrective maintenance

activities involving the operators.

2. The load design data sheets (Drawing E1012, Revision 4,'" Motor

Operated Valve Data"), specifying the required torque switch

settings, were incomplete. The inspector reviewed the data for

160 Limitorque operated valves and found five instances of no

specified torque switch setting (HV-20003, HV-23801, HV-23802,

LV-36005, LV-36505). Three of these examples involved Class 1

-

valves. This. item is unresolved, pending further review of the

. safety function of the valve operator. (86-21-07)

~3. The inspector noted that the licensee's approach to setting the

torque switches involved application of a measured torque to

the handwheel of the operator. As described in the May 16,

1986, submittal of their MOV program,Section III.4, "the

torque switch settings will be input through the actuator

handwheel and set to a specific value of handwheel torque."

'_

However, the inspector observed that the manual mode of

operation of the smaller style Limitorque operators, SMB-000

and SMB-00, does not actuate the torque switches.

Consequently, the torque switches of these operators cannot be

adjusted based on an applied handwheel torque. In this

respect, the licensee's. implementation of'their MOV program

appears questionable. The inspector determined that 24 of the

30 valve operators to be inspected were the smaller style

l, Limitorque operator which do not actuate the torque switches in

'

the manual mode.

,

4. Section III.4 of the licensee's MOV program describes that "the

open torque bypass switch setting will be'a fixed percentage of

,

- -- ,. y . . _ , . . - . . , _ . , - . , , , , . . _ ~ - 1.,...,_,-._ , _ . - , , - - . _ . _ _ . , w.. _r,. ..

i

+- i

dise (emphasis 'added) motion for all motor operated valves".

The inspector noted the licensee's current procedure EM.117,

Revision 5, specifies setting the torque bypass switch "to open

as the valve stem (emphasis added) lif ts off the valve seat.

This position may be determined by noting a marked reduction in

the force required to manually drive the stem open. If, for

some reason, this condition is not detectable, the torque

-

switch bypass should be set to open at approximately 10 percent

of total valve travel". The inspector noted that EM.117 did

not address valve disc motion as committed in Section III.4

since valve disc'and stem motion may not be equivalent. In

discussion with-licensee maintenance personnel,-the inspector

also determined that the torque bypass switch-was routinely set

at 10 percent of total handwheel' turns to drive the valve stem

, open, which is not necessarily equivalent with 10 percent of

total valve travel (or disc motion).

The inspector discussed these concerns regarding the technical adequacy

of the MOV program with licensee' representatives, who stated that the

inspector's concerns would be addressed in the analytical determination

of the valve stem thrust and an extensive revision of the EM.117 which is

'

currently underway.

,

This bulletin followup will remain open.

10. Meetings

On June 17 and 18, 1986, members of the NRC Regional and Headquarters

office met with licensee representatives at the Rancho Seco site to

-discuss the Rancho Seco Improvement Program. The licensee's presentation

included discussions of the Program's various input phases (personnel

interviews, precursor review, B&W Stop Trip Program, selected projects,

deterministic failure analysis, and NUREG 1195); the responsibilities of

the various committees (Recommendations Review and Resolution Board,

Performance Analysis Group, Independent Analysis Group); and the

structure of the system test program. The meeting was an information

meeting only.

Also, the licensee issued their Rancho Seco " Action Plan for Performance

Improvement" on July 3, 1986.

No violations or deviations were identified.

'

11. Personnel Changes

During this report period various changes have been made in the site and

corporate managements. General Manager D. K. K. Lowe resigned after ten

months, in his place the Board of Directors appointed William Latham.

William Latham has been with SMUD for 25 years and most recently held the

Assistant General Manager, Consumers Services, position.

The following changes have been mede with the site management: John Ward

(Management Analysis Corporation, MAC) was named the Deputy General

Manager, Nuclear; Dan Poole (MAC) was named Restart & Implementation

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_ _ __ ,

.. .

.

. .

14

^ Manager; Ray Ashley (MAC) was named Manager, Nuclear Licensing; and

Stu Knight (MAC) was named site operations Quality Department Manager.

12. Unresolved Items

An unresolved item is a matter about which more information is required

in order to ascertain whether it is an acceptable item, and open item, a

deviation, or a violation.

13. Exit Meeting

The resident inspectors met with license'e representatives (noted in

Paragraph 1) at various times during the report period and formally on

July 11, 1986. The scope and findings of the inspection activities

described in this report:were summarized at the meeting.

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