ML20206R705

Insp Rept 50-312/86-18 on 860422-0531.Violation Noted:Review of Mod to Reactor Coolant Pressure Not Performed Under Nonconformance Rept 5218 & Abnormal Tags on Temporary Mods

ML20206R705
Person / Time
Site:	Rancho Seco
Issue date:	06/20/1986
From:	Albert W, Miller L, Myers C, Perez G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20206R623	List: ML20206R615 ML20206R635 ML20206R673 ML20206R705
References
50-312-86-18, NUDOCS 8607070208
Download: ML20206R705 (13)

See also: IR 05000312/1986018

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No: 50-312/86-18

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 conduct d

Inspectors: N '

G. P'. P re z , S e r Repident Inspector (Acting) Date Signed

A (o - ?p %

, C. J. Olyerrs, Res , ntlInspector Date Signed

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s , A G-20-26

W. Alberg, Regi Inspector Date Signed

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L. F. y ller, Chief Date Signed

Reactor ProjectsSection II

Summary:

Inspection between-April 22 to May 31, 1986 (Report 50-312/86-18).

Areas Inspected: This routine inspection by the resident inspectors and by a

regional inspector, involved operational safety verification, maintenance,

surveillance, and followup items. During this inspection,. Inspection

Procedures 30702, 30703, 36700, 42700, 61726, 62703, 62700, 71707,

92700, 92701, 92702, 93702 and 94703 were used.

Results: Of the areas inspected, two violations and one deviation were

identified. One violation was in the area of PRC cafety reviews, and the

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.other violation involves abnormal tags on temporary modifications. The

deviation involved a procedural change commitment from a previoris violation.

In addition, two unresolved items were identified in the area of piping

, classification, and a third unresolved item involved 10CFR50.59 review

requirements.

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DETAILS

1. Persons Contacted

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a. Licensee Personnel

. *G. Coward, Manager,' Nuclear Plant

J. McColligan, Assistant Manager, Nuclear Plant

• S. Redeker, Nuclear Operations Manager

• M. Price, Nuclear Mechanical Maintenance Superintendent

• R. Colombo, Regulatory Compliance Superintendent

! J. Field, Nuclear Technical Support Superintendent

S. Crunk, Incident Analysis Group Supervisor

• F. Kellie, Acting Chemical and Radiation Supervisor

• L. Schwieger, Quality Assurance Manager

J. Jewett, Site QA Supervisor

H. Canter, QA Operations Surveillence Supervisor

• C Stephenson, Regulatory Compliance Engineer

B. Daniels, Supervisor, Electrical Engineering

D. Army,' Nuclear Maintenance Manager

B. Croleyi Nuclear Technical Manager

• M. Cooper, Nuclear Operations

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C. Linkhart, Electrical Maintenance Superintendent

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• T. Khan, Principal Mechanical Engineer

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• D. Gillispie, Nuclear Engineering Manager

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.0ther licensee employees contacted included technicians, operators,

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mechanics, security and office personnel.

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• Attended the Exit Meeting on June 3, 1986

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2. . Operational Safety Verification

The plant was in cold shutdown. condition during the entire period of this

report. .The plant has'been in this mode of. operation since the

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December 26,1986 plant trip initiated by a loss of integrated control-

system power.

During this period the inspectors' observed control room operations,

verified proper control room stnffing, reviewed applicable logs,

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l conducted discussions with the operations crews,= reviewed selected

, emergency systems,Lreviewed tag-out records, verified proper removal of,

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affected components.from service, and verified the licensee's adherence-

to limiting conditions for oparatione.

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Tours of the auxiliary, turbine, and reactor buildings and'the general

site area were (2nducted to observe plant equipment condition,'and to

1 . verify,that maintenance requests had been initiated for equipment =in need

!- of maintenance.-

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The inspectors reviewed portions of.no.t-license'd operator logs, conducted

various discussions with the non-licensed operators and observed:them-

performing their assigned duties. ,

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During tours of the facility, the inspectors entered radiologically

controlled' areas. The inspectors verified compliance with the licensee's

radiation protection program. The inspectors discussed the radiation

work permit requirements and the radiological conditions of the work

areas with workers in the radiologically controlled areas. Also, the

inspectors . verified proper clothing requirements and observed the method

of personal frisking when exiting radiological controlled areas. The

inspectors examined selected radiation protection instruments to verify

their. operability and calibration.

The licensee's adherence to the physical security plan was evaluated

daily during this period by observing the entry process, wearing of photo

identification badges by personnel, escorting of visitors, and security

compensatory measures.

No violations or deviations were identified.

3. Monthly Maintenance Observation

Several maintenance activities for the systems and components listed

below were obcerved 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

removed 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

Laplemented; and fire prevention controls were implemented.

A. Temporary Modifications

The inspector reviewed Nonconformance Report (NCR) #5218 dated

December 8, 1985, which dealt with a repair of the inlet isolation

valve (SFV-22006) to the B letdown cooler. While the plant was in a

hot shutdown condition, the Class I, 2 1/2" valve had developed an

external leak through the gasket at the body-to-bonnet joint.

NCR #5218 was dispositioned to temporarily patch the leak by using

an approved sealant, and the valve was returned to service until

rework of the gasket seating surfaces could be performed under cold

shutdown conditions. This temporary repair involved the

installation of a 70 lb. external clamping ring around the

body-to-bonnet joint to serve as a retainer for the sealant. The

valve was returned to service with the clamping ring attached after

the sealant was injected.

The inspector found'that vben the NCR had been originally screened

for applicability under the facility modification review

requirements of 10CFR50.59, it had been checked "10CFR50.59 review -

No". This screening had been performed prior to dispositioning of

the discrepant condition.  ;

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The inspector reviewed Section 4.3 of the Updated Safety Analysis

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Report.(USAR) and found that the dead weight loads applied to the

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reactor coolant' system valves were described as part of the design

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{ USAR Section-4.3.5 states in part, "To assure the integrity of the

system during a seismic event, it was specified that reactor coolant

system valv'es would be capable of sustaining 3.6g loading without

exceeding the stresses specified in the code. - Furthermore, the

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specification supplied specific information regarding the thermal

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dead weight,and seismic loads imposed upon the valves.by the,

connecting piping to be accomodated in the valve by the valve-

manufact'urer". Section 4.3.9.4 states, in part "The valve vendor

was also required to submit calculations showing that when the valve

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assembly is subjected to a 3g horizantal force and a 2g vertical

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force, stresses incurred are within.the code allowable stresses.

The District has also verified that the first natural frequency is

above 20 cps..." r

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j The structural effects on the piping system.due to the additional

weight of the clamping ring were evaluated by engineering as part of

the NCR disposition and found to be negligible.

However, the inspector reviewed the engineering calculations and

found that no specific review of the effect of the additional weight

on the previously analyzed natural frequency of the valve had been

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Also, as a result of the NCR designation as "50.59 'No", no: safety

evaluation had been performed to determine if the temporary

modification affected the previous analysis described in the USAR or  !

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whether it involved an unreviewed safety question.

The inspector determined.that the Manager, Nuclear Plant or his

designated alternate performed the initial screening of NCRs to

determine if a safety analysis was required to meet the requirements

{ of either 10 CFR 50.59 and Technical Specification 6.5.1.6d. g

j However, no specific screening criteria for evaluating the j

discrepant condition reported-in the NCR were part of the licensee -l

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' program described in QAP-2, Design Control, or QCI-5, Safety Review

of Proposed Facility Changes.

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Furthermore, the initial 10 CFR 50.59 screening was' performed prior

to dispositioning of the NCR. This-precluded evaluation of the i

corrective actions specified in terms of the 10 CFR 50.59

i requirements unless a subsequent Engineering Change Notice'(ECN) wcs

j ' generated to accomplish the wo'rk. The temporary modification

included as part of the' disposition of NCR #5218 was performed under

! a work. request and was, therefore, never reviewed'for.10 CFR 50.59

cpplicability.

This item is unresolved pending a determination of whether the

modification reduced the valves natural frequency below 20 Hz.

} (Unresolved Item 86-18-01)

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-Rancho Seco Technical Specification 6.5.1.6(d) requires that the

Plant Review Committee (PRC) review all proposed changes or

-modifications to plant systems or equipment that affect nuclear

safety. Quality Control Instruction QCI-5 defined systems or

equipment that affect nuclear safety as including those necessary to

assure the integrity of the reactor coolant pressure boundary.

The licensee was unable to retreive any record of PRC review of the

temporary modification to valve SFV-22006 made under NCR #5218.

This failure of the PRC to review the temporary modification of the

reactor coolant pressure boundary valve SFV-22006 is an apparent

violation of the Technical Specification 6.5.1.6(d). (Enforcement-

Item 86-18-02).

The inspector further identified a weakness in the licensee's NCR

procedure QCI-5, which did not incorporate a specific method to

designate whether a PRC safety review was required. The procedure

only identified items for PRC review if the activity involved a

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50.59 change. If it was determined.that a 50.59 review was not

required during initial screening of the NCR, no further

determination was made as to whether the PRC should review the

change if it may affect nuclear safety as required by technical

specifications.

The inspector reviewed the licensee's abnormal tag procedure,

AP.22 (Rev.ll) which controls and documents temporary changes to the

facility. Thetinspector found that an abnormal tag was required any

time an electrical, mechanical, structural or pneumatic system was

modified and placed in service without an approved drawing change

notice (DCN) per NEP 4109, to document the change. .The significance

of this item was that the inspector found that 10 CFR 50.59 screening

was performed as part of AP.22 and that the PRC reviewed all Class I

equipment abnormal tags within one working day.after the tag is

hung. In this case, however, the temporary modification to valve

SFV-22006 was not controlled under the licensee's approved procedure,

AP.22. This is an apparent violation. (Enforcement Item 86-18-03).

B. Weld Rod Control

The inspector observed portions of the work activity in progress

under Work Request 114810 to install new battery racks for the new

safety related station batteries,

While observing welding on the battery racks, the inspector noted a

weld rod heater can containing low hydrogen weld rods (type 7018)

, which was not plugged icto at electrical outlet. Both the heater

can and the weld rods were cold. The inspector brought this

condition to the attention of the welders and quality control

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personnel for resolution.

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The licensee's procedure for control of welding filler material,

QCI-109, paragraph 2.4-5, required that the portable rod ovens shall

be continuously heated except during transit.

The licensee documented the inspector's observation in an Occurrence

Description Report (ODR #220), and the welder immediately corrected

the condition by bending the subject weld rods and returning the

heater can to the tool room where the weld rod was accounted for and

disposed. The licensee further identified that no welding had been

performed using any of the cold weld rod. In addition, the licensee

determined that no significant hydrogen contamination of the weld

rod would have occurred in any case for the short period of time

that the weld rod was unheated to result in any adverse effect on

the weldments being made.

The licensee determined that the two welders had consolidated their

weld rods, which were both of the same type, into another heater can

due to a lack of electrical outlets in the area. They left the

heater can (observed by the inspector) with extra weld rods

intentionally unplugged prior to returning the can to the tool room.

The inspector and the licensee agreed that the intent of the

administrative control of weld filler material was to ensure that

each welder only use weld rods that were issued specifically to him

for a particular job. A licensee representative stated that the

procedural intent would be clarified in the next revision of their

procedure. (0 pen Item 86-18-04).

4. Monthly Surveillance Observation

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: -testing 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 directirg 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.

The following surveillances were reviewed:

STP-958 Station Battery Service Test - SFAS Load Profile Discharge

Test

STP-198 Control Room / Technical Support Center Essential Air

Filtration Unit Flow Test.

No violations or deviations were identified.

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5. High Point' Vent Piping Classification 'I

During followup of the June. 23,-1985 event in which a crack in the

reactor coolant system high point piping resulted in a non-isolable

reactor coolant leak, the NRC began an assessment of the appropriate

piping classification for this system. .,

In a letter dated March 4, 1986, NRR provided a safety evaluation report

(SER) which addressed (1) the consequences of an unisolable one-inch

diameter, pipe break and (2). interpretation of regulatory._ requirements

pertaining to the quality classificaiton of the high point vent system.

The SER concluded:

(1) A postulated one-inch diameter pipe break during operating

conditions would be classified as a small break loss of

coolant accident (SBLOCA) since the break flow was beyond the

single makeup pump capacity.

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! (2) The high point vent piping up to and including the second

shutoff valve should be classified as Quality Class I, Seismic

Class I and constructed to ASME, Section.III, Class I (or

j USAS B31.7 Class I) requirements.

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(3) The nitrogen supply line up to the pipe supports performs a ,

safety related function because the nitrogen supply line

supports also support the high point vent line. The portion of

the. nitrogen supply system-up to the supports beyond the second

shutoff valve should be classified Seismic Category I, Quality

Assurance Class I, and constructed to ASME III, Class 3 (or

USAS-B.31.7, Class 3)..

Licensee representatives stated that the current piping classification of

the high point vent was consistent with their code commitments affecting

, the reactor. coolant pressure boundary at the time of original

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construction. As such, in their view,"the Quality Class I,' Seismic

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Class I classification properly' extended only to the first isolation

valve. ,

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The apparently improper classification of the vent piping is an

unresolved item. .(Unresolved Item 86-18-05).

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-The inspector. discussed the SER with licensee representatives to clarify

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"the basis for the existing vent pipe classifications.. . Licensee

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representatives acknowledged that the current piping classifications did

not identify the extent of seismically analyzed piping beyond the

designated Class I boundary. The licensee is developing a program that

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will identify.and control the-piping and supports of non-seismic Class I 4

systems that are used in the seismic' analysis.

This item is considered to be unresolved. (Unresolved Item 86-18-06).

At the exit meeting, the licensee, committed to submit a written

explanation'and-justification for the high point vent piping-

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classification. This submittal will also address the licensee's plans to

institute programatic changes to administratively control the. piping and

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supports of non-Class I systems that are used in the seismic analysis.

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6. Restart' Issues

i Duringthisreportperiokthefollowingrestartissues,whichwereamong

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those addressed in Report 50-512/86-07, were further examined. Each

l . restart issue has a Region V designator and some have one or more Open

l Item List ~ (0.I.? List) designations,

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j A. -(Closed)' Restart Issue E-7,. Complete Post Trip Report-

'Regarding the. transient of December 26, 1985, the licensee provided [

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the NRC with'the report prepared for this reactor trip. The

transient of December 26, 1985 wasttrip No. 75. The report was .

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revised on April 15, 1986 and was received by the NRC on- _

! May 21,1 1986 after.it had been reviewed by the licensee's onsite

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safety committee. The inspector concluded that most issues raised

j in the report were being adequately addressed. One possible

exception was the question of floor drains from the Control Room and

Technical Support Center (TSC). The report pointed out that no

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floor drains were provided in the rooms, despite being protected by

a sprinkler system which alarmed -(but not actuated) at the _ time of

l the 12/26 event. The potential for serious flooding of the rooms,

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should the system actuate was noted by the inspector.

The licensee addressed the control of the small amount of water

vhich is released when the sytem alarms, such as that which occurred

i during the 12/26 event. However, water control by the licensee.if

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the sytem were fully actuated has not been addressed yet by the

! licensee. This issue was discussed with the' licensee a'd n will be

j further examined during a future' inspection. (0 pen Item 86-18-07).

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j The restart issue of completing a post trip report, Item E-7 on the

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Region V restart list is CLOSED.-

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} B. (OPEN) Restart Issue MA-6, Determine Disposition o'f Degraded' 125 V

j Station Batteries (See also 0.1. List items 50-512/86-07-12 through

j 17, all OPEN).

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1. General

A meeticg was held with the licensee to cetermine. progress ot.

the various parts of this issue still to be resolved..

. The licensee first discussed the history of the battery

, problen. SMUD pointed out that.the degraded condition was-

! first noted by the licensee's service personnel in i

November 1965 and arrangements.to have the~ batteries examined "

by a manufacturer's. representative were made at that time. -The

j manufacturer's representative confirmed that the condition of ,

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the batteries was such that replacement was required.'.This'

determination was made in February 1986.

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. . 2. Seismic' Qualification of Battery Racks (86-07-17)

The recognition that the station battery racks were_not in the ^

seismic qualification configuration was first noted when the

NRC' examined the licensee's findings regarding the degraded

condition of the batteries. The licensee stated that when they

had received a letter from the manufacturer (Gould)lin

March 1985 regarding seismic criteria, they inspected their

recent new battery installation.in the Nuclear Services

Building but failed to assertain that the original station

batteries met the Gould criteria.

The inspector observed the work underway for the installation

of the new battery racks. The workmanship appeared good.

j However, one concern was identified in the area of weld rod

control, this is addressed in paragraph #3 in this-report.

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1 This item remains open. '

3. Capacity Testing (86-07-16)- l

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The licensee stated during discussions of capacity _ testing that

it would be done on a five year basis in the future. The ,

licensee nonetheless did not believe that a definitive

requirement existed for such tests in any industry standard.

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j Until resolved by NRR this matter remains open.

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4. Service Testing (86-07-14)

The licensee stated that the expressed _NRC concern about the

i wording of procedure EM-106 did not imply that service tests

were being conducted following immediately after an equalizing  ;

charge of the batteries.. However, they agreed that the wording-

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of'EM-106 could be misleading and the. procedure is being

q changed to clarify this matter. This item will remain open. .

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5. Weekly Battery Surveillance Program (86-07-13) ,

The licensee stated that the practice of using the worst cell

i as a pilot cell will be discontinued.. Also the two errors . *

j noted by the NRC.with regard to Battery Pilot Cell Tests had.

been investigated by the licensee. The licensee provided data

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establishing that one of the two apparent failuren to' equalize-

bank BD (June 17, 1985) had actually been accomplished.- This

item will remain'open. "

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6. Schedule for Installation

The replacement. of the batteries is L currently expected to be

accomplished by the end of July 1986. The licensee. expects to

] utilize 60 cells per bank:as in _the original design-but will-

later perform the -engineering necessary to determine if it: may

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be desireable to use 58 or 59 cells.as has been done at certain1

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other nuclear facilities.

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7. Organization and Administration

The inspector reviewed portions of Sectio'n"6.0 " Administrative Controls"

of the licensee's Technical Specifications (T.S.) against the licensee's '

present organization. The inspector identified the following

T.S.

Section 6.1.1 states that the Plant Superintendent shall be responsible

for the operation and maintenance of the plant.- At;the end of this *:

, report period the Nuclear Operations Manager had the responsibility for

1 plant operations and the Nuclear Maintenance Manager had the-

l responsibility for plant maintenance. In addition, the Facility

l organization charts figures 6.2-1 and 6.2-2 did not : reflect the current

j organization structure,

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!' These items were brought to the licensee's attention. A licensee's ,

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representative stated that although'the titles of the position's have

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l changed under the' plant's current organization, there is a one for one- '

! correlation of_ responsibilities from the T.S. charts to the current

charts. The inspector also found that the qualifications required by;

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ANSI N18.1;1971 for the appropriate supervision levels are currently met.

The licensee had sent a proposed amendment to change the organizational

> charts in the T.S. in December 1985. This amendment has not,yet_been ,

issued by the NRC, therefore the licensee has committed to submit a

revised amendment addressing the current facility organization by

August 1, 1986. The inspector found that_these actions will resolve the-

conflict between the T.S. and the present organization. This item '

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remains open., (OI 86-18-09)

i 8. Followup Items'  ;

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(CLOSED) Violation 85-30-01 "Cooldown Rate Exceeded Technical

! Specification Limit" .

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During the October 2, 1985 event, due to steam being relieved *

3 through feedwater. heater relief valves, the plant experienced 'a

cooldown which exceeded the technical specification limit. To  ;

i correct the cause of the cooldown the licensee raised the relief

! valve setpoints and reduced the pegging steam' pressure controller "

! setpoints. This was to elininate an overlap in the setroints. L

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After this adjustmenti no integrated system test of the pegging

steam controller and the relief valve was performed. On f

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December 5,.1985.a reactor trip occurred and the relief valves again ,

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lifted. Testing following.this occurrence lfeund.that the steam-

controller had an overshoot characteristic which was not identified ,

i previously. The licensee evaluated the finding,:further reduced the,  ;

{ controller setpoint pressure. and successfully performed an  !

integrated surveillance test on the pegging steam system.

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! Therefore, the inspector concluded that because the. licensee

performed a,nore thorough _ test after the December 5, 1985 reactor

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! trip, .the adjustments to the controllers and relief valves should -

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preclude excessive steam flow from the pegging steam system with the

consequent cooldown, on a reactor trip.

In addition, on December 26, 1985, a cooldown occurred, caused by

events different from the above described, but the cooldown was more

excessive. The licensee has addressed the effects of the

December 26, 1985 cooldown and has stated that the cooldown was

within the parameters of the design for the reactor vessel. The NRC

is reviewing the effects of the cooldowns to the reactor vessel.

This item is closed (85-30-01).

(CLOSED) Violations 83-25-01/02/03/04/05

Inspection Report 83-25 identified five violations which involved

the breakdown in communications between and within organizations,

failure to adhere to procedures, and failure to properly notify the

regional office of an event within the technical specification

(T.S.) time limits. Because these violations were based on a

singular event, the licensee's corrective actions for all the

violations will be addressed here.

The inspector verified, by personnel interviews, that the licensee

has trained key personnel on the identification and reporting

requirements of events. The licensee also revised the chemistry

manual to provide more direction when an out of specification sample

is identified, This modification provided procedural steps in

communication within the organization. The reporting of the event

to the NRC was not accomplished per the requirements; in the period

since the event the NRC reporting requirements have significantly

changed, and so have the requirements in the T.S. Presently, only a

thirty-day LER would be required for this event. The inspector

found that the licensee had generally been reporting required LERs

on time. The licensee also stressed the importance of procedural

adherence, and has provided memos stating this position to the

licensee staff. Therefore, these five enforcement items are closed

due to the efforts the licensee has taken to correct the problem and

their corrective actions to avoid further occurrence. Items

83-25-01 83-25-02, 83-25-03, 83-25-04 and 83-25-05 are closed.

(CLOSED) Violation 65-30-02 - Use of an unapproved procedure

The violation, in summary, wac a result of che licersee using an

unapproved procedure to change the gain on a nuclear instrument.

The inspector's review of the licensee'a response to violation

85-30-02 found it to be narrow in scope. No actions were detailed

by the licensee on the implications of personnal performing work

without approved procedures. The correceive ection taken by the

licensee to prevent reoccurrence, to some extent, should

involve more than just the revision of one procedure. It should

also be broad enough to assure management other unapproved

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procedures are not used. Therefore, the inspector requested the

licensee to amplify their response to include measures to:

prevent the reoccurrence of the use of any unapproved

procedure

prevent the failure to satisy their commitment

A licensee representatise agreed to address these areas in their

response to the the deviation identified in this section.

The licensee, in a letter dated December 23, 1985, had only

committed to incorporate a temporary change to procedure I.103 and

then to make a permanent revision to procedure I.103 by

February 28, 1986. At Rancho Seco, temporary procedure changes are

routinely voided after a defined internal (typically 30 days). In

this case, the temporary change expired in October, 1985. -The

inspector was able to identify that the temporary change had been

made to the procedure, but was not able to find a permanent revision

to procedure I.103. Therefore, the failure to satisfy a commitment

detailed in the response to violation 85-30-02 is considered an

apparent deviation. (Deviation Item 86-18-08)

The inspector brought the above item to.the Maintenance Manager's

attention and discussed the importance of satisfying commitments.

In addition, Item 85-30-02 is CLOSED, deviation 86-18-08 is OPEN.

(CLOSED) Followup Item 85-04-03 Inaccurate reporting of piping system

classification in LER 85-01

In Inspection Report 85-04, LER 85-01 was reviewed. The inspector

found that the seismic and quality classification of the hydrogen

monitor was incorrectly reported in the LER. Tids was discussed in

the report and with the licensee at the inspector's exit. The

licensee committed to revise the LER and include the proper

classification.

The licensee submitted revision 1 to LER 85-01 on April 28, 1986.

The inspector found that the licensee corrected the classification

of the piping system. Thus, this item is closed.

Follosup of Licensee Event Report (LER) (CLOSED) (85-23) " Reactor Trip

on High Pressure Resulting From Stean Generator Underfeed" l

The inspector followed up on the LER 85-23 to ascertain whether the

licensee's review, corrective action and reporting of the identified

event and associated conditions were adequate.

The inspector reviewed the LER and found the description adequately

reflected the trip and the corrective actions taken to prevent

reoccurrence. The trip, on December 5, 1985, has already been

discussed during routine followup of events in inspection report

85-32. This report also details the review of the licensee's

corrective actions.

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After the plant trip, the pegging steam relief valve on the 4A

heater lifted due to an overshoot in the pegging steam supply

controller. -This relief valve had also lifted after the

October 2, 1985 trip, and the licensee took corrective actions in

readjusting the valve's set point. It is clear that the corrective

actions and post modification testing done to the relief valve after

the October 2, 1985 event was incomplete. It wasn't until the

December 5, 1985 trip and subsequent integrated system testing of

the pegging steam system that the overshoot problem was identified.

Therefore, the events that occurred af ter the December 5,1985 trip

identify a need to test the whole system after a modifications in an

integrated manner, even though only one portion of the system was

directly effected by the modification.

In addition, the licensee committed to completing a review of this

event through their root cause system. The root cause report was

issued on April 21, 1986.

Based on review cf the LER and verification of the licensee's

corrective actions this item is closed.

9. Commissioner Meetings

During this report period the inspectors were involved with two visits

with NRC Commissioners at the Rancho Seco site. The first visit was with

Commissioner Bernthal and his assistant on April 24, 1986. The resident '

inspector gave a tour which involved the control room, battery room,

turbine building, auxiliary building, tank farm, auxiliary feedwater pump

area, and the valves that had failed to operate properly during the

December 26, 1985 event. A licensee briefing of the December 26, 1985

event and the program initiated to restart the plant were presented..

After the briefing a short press conference was held with the

Commissioner, and he then'left the facility.

On May 14, 1986 Commissioner Asselstine and his assistant visited the

site for a similiar tour as mentioned above. In addition, the tour with

Commissioner Asselstine included a tour of inside the reactor building.

This visit was also followed by a press conference.

No violations or deviations were identified.

10. Exit Meeting

The resident inspector met with licensee representatives (denoted in

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

June 3, 1986. The scope and findings of the inspection activities as

given in this report were summarized at the meeting.

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