ML20215C337

Insp Rept 50-312/87-16 on 870504-08.Violations Noted: Inadequate Inservice Testing Procedures for Valve Full Stroke Testing

ML20215C337
Person / Time
Site:	Rancho Seco
Issue date:	05/27/1987
From:	Ang W, Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20215C304	List: ML20215C303 ML20215C324 ML20215C337
References
50-312-87-16, NUDOCS 8706180080
Download: ML20215C337 (22)

See also: IR 05000312/1987016

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U. S. NUCLEAR REGULATORY COMMISSION

, REGION V ,

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Report No. 50-312/87-16 .'

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Docket'No. 50-312 '

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License No. 'DPR-54-

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Licensee: Sacramento Municipal Utility' District (SMUD)

P. O. Box 15830

Sacramento, California 95813

Facility Name: Rancho Seco Nuclear Generating Station 1

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Inspection Conducted: May 4 to'May 8, 1987

Inspected by: . /A) y S-2n- t 1

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W. P ,.Projs - spector Date Signed

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Approved by:

L lier, Chief, Pr6 ject Section 2

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Date Signed

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. Accompanying' Personnel: D.' A. Beckman, Prisuta-Beckman Associates

~ Summary:.

Inspection on May 4 to May 8, 1987 (Report No. 50-312/87-16)'

Areas Inspected: Routine announced inspection by a region base'd inspector and l

, an NRC contractor of licensee action on previously identified inspector items,

Licensee Event Reports, and personnel qualifications. Inspection Procedures

30703, 36100, 92700, 92701, and 92702 were'used during this inspection.

Results': In the areas inspected, one violation regarding inadequate inservice I

testing procedures for valve full stroke testing (paragraph 4. A.) was

identified.

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8706180000 870529

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DETAILS , ,

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F < ll ' Personnel Contacted y

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• B. Croley, Plant Manager.-

• S. Knight, QA Manager

<- W.,Kemper, Operations Manager.

D. Army, Nuclear Maintenance Manager' ,

'J. Vinquist, Licensing Manager (Acting) , ,

G. Shanker, Deputy Maintenance Manager,

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G. Clefton, Assistant Maintenance Manager. '

*D. Brock,' Maintenance Programs Supervisorf . t

• S.'Redeker, Assistant to the Operations Manager.

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D. Wiles, I&C Supervisor

H. Humphrey, I&C Supervisor

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J. Lingenfelter, Systems Review and Test Program Assistant Director

'*J. Robertson, Nuclear Licensing Engineer

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• A. D'Angelo,. Senior NRC Resident Inspect'or

• Attended the exit meetings.

'The inspector also held discussions with other licensee and contract

personne1'during the inspection. This included plant staff engineers,

technicians, administrative and clerical assistants.

2. Licensee Action on' Previous Enforcement Matters

A. (Closed) Violation 86-18-02, Failure of PRC to Review Temporary

. Valve. Modification per TS 6.5.1.6(d), and ..

(Closed) Violation 86-18-03, Failure to1 Implement' Abnormal Tag

Procedure Requirements for Temporary Valve Modification

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Valve SFV-22006 was modified by installation .of a temporary clamping

ring and sealant to correct a body to bonnet reactor coolant. '

pressure boundary leak (see Unresolved Item 86-18-01, paragraph 4.B

for bect. ground information).

Technical Specification 6.5.1.6(d) requires that the Plant Review l

Committee (PRC) review all proposed changes or modification to plant

systems or equipment that affect nuclear safety.. Violation 85-18-02

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cited the failure of the PRC to review the above andification.

Rancho Seco Administrative Procedure AP-26, Abnormal Tag Procedure, i

requires an Abnoratal Tag be issued any time a system is modified and

placed .in service without an approved Drawing Change Notice.  !

Violation 86-18-03 was issued for failure to process an Abnormal Tag .

for the modification. I

The licensee provided a combined response (JEW 86-896,~ November 20,

'1986). for the two . violations, addressing five commitments / actions as

discussed below: 1

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N The licensee committed to having the PRC review all temporary

changes to.any Class.1 systems.or components in the plant,  ;

regardless'of the Part.50.59 determination. '

This! commitment is: implemented by revision 13 to AP-26,

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Abnormal. Tag Procedure, which now requires that all' Abnormal

' Tags be provided to and reviewed by the PRC.within one working

day. . Inspector's review of the Abnormal. Tag log confirmed.

implementation.  !

The Nonconformance' Report (NCR) Procedure (QAP-17) and Abnormal ,

Tag: Procedure.(AP-26) were revised in accordance with the

1icensee's response to include' provisions to invoke the

Abnormal = Tag process.and 10 CFR 50.59 reviews. Revision 4 of

.QAP-17 included addition of a '" Note" requiring implementation

of AP-26 if temporary repairs or interim accept-as-is"

dispositions'are specified for an NCR. AP-26, revision 13,

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incorporated a new'10 CFR 50.59 evaluation questionnaire and

, provisions for.PRC review as previously noted above.

,The licensee's response noted that.both' procedure revisions

vpuld be made byL January 5,1987. . However, revision 4'of-

QAP-17.was not approved and issued for' distribution until

Jhnuary 11, 1987; revision 13 of AP-26'was not effective until

March'30, 1987. This failure to meet commitment dates to NRC

is essentially. identical to Deviation 87-11-01 issued on

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April.-20,.1987; the licensee's. response to Deviation 87-11-01

was p6nding at the. time of this inspection. The licensee was

' advised that-their response to Deviation 87-11-01 would be

evaluated.in light of the-apparent-recurring nature of the

problems described therein and above.

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The licensee's response to this item further committed to

conduct a' review of the. technical adequacy of previous

temporary changes to Class 1 systems implemented by NCRs by i

February 2, 1987. Where establishment of adequacy was not

apparent,.an_ assessment would be performed. 1

SMUD memorandum QLC 87-006, dated January 26, 1987, documented

the results of the QA/QC review of all open Class 1, 2, and 3

NCRs through December 31, 1986. A total of 747 NCRs were ,

reviewed; three NCRs were identified as having implemented i

temporary changes without performance-of 10 CFR 50.59 reviews.

The three NCRs were provided to the PRC for review and the PRC

review of the NCR's for compliance with 10CFR 50.59 were ,

documented in PRC Meeting Minutes No.1637, January 30, 1987.

The inspector reviewed the NCRs and the PRC review conclusions

and found the PRC review to be acceptable.

The licensee review also included a statistical sample of 260 l

out of 4,000 NCRs issued during the period July 1980 through

December 1986. The' review found that none of the samples

involved NCRs implementing temporary changes to the facility

that were not subject to 10 CFR 50.59 and Technical Specification

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PRC reviews, concluding that the problems ideritified by NRC

Inspection 50-312/86-18 were a recently developed trend.

The licensee also committed'to a programmatic review of the i

District's procedures for the control of modifications pursuant

to 10 CFR 50.59 by December 8, 1986.' Memorandum QLC 86-121 -l i

issued Audit Report No. 0-846, " Response to Inspection 86-18,"

on October 30, 1986, for the audit completed on September 25,

~1986.

The audit recommended: revision of existing procedures to. '

clarify and consolidate programmatic requirements; provision of j

additional staff training; incorporation of INP0 Good Practice

TS-415 provisions;' application of 10 CFR 50.59 provisions to

changes to Nuclear Engineering Procedures; and, others. ]

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A new procedure, RSAP-0901, " Safety Review of Proposed Changes,

Tests, and Experiments," Draft, had been issued to_ replace the

combined requirements of QCI-5 and QCI-18; to incorporate the

Audit 0-846 findings and recommendations, and meet new licensee

procedure upgrade program requirements. The procedure had been ,

PRC and MSRC approved but not yet implemented pending personnel

training.

Although this inspection confirmed that the licensee's corrective

and preventive actions for Violations 86-18-02 and -03 had been

implemented, NRC Inspection 50-312/87-06 identified two similar

violations (87-06-01 and -02) wherein temporary repairs had been

made to a leaking pipe in the Nuclear Service Raw Water (NSRW)

system without issuance of a Nonconformance Report, Work Request ,

and/or Abnormal Tag for a temporary patch. Violations 86-18-02 and 4

-03 will be closed for record purposes. Apparent continuing

weaknesses in the licensee's programs for timely completion of '

commitments to NRC and in programs for control .of temporary

modifications will be reviewed upon NRC receipt of the SMUD

responses to NRC Inspections 87-06 and 87-11.

3. Licensee Action on Region V Open Items Regarding Inspection Report

50-312/86-07 and the December 26, 1985 Event

A. RV - MA-4 (0 pen)

SMUD - 16c(4)

EDO - 4a

Verify Operability of Manual Vaives and Remote Operated Valves.

Perform Inspections to Assure Irategrity of Packing and Verify Proper

Assembly of Manual Operators Including Setting of " Neutral" Position

and Mounting Devices

Inspection Report 50-312/86-07 and 87-11 documented NRC inspection

of the licensee's inspection, repair and testing of manual and

remote operated valves. Inspector's concerns regarding the manual

valve inspection and preventative maintenance program were

identified in inspection report 50-312/87-11 as follows:

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(1) The licensee did not appear to have a reviewed and approved

document that lists all manual valves which are critical for

plant operations. The list of 142 valves appears to have been

developed by an informal piece-meal process, rather than by a

formal, pre-planned review process. The NRC considered that

the importance of this activity should have been subjected to

a review and approval by appropriate levels of management.-

(2) The inspections of the valves appear to have been performed by

maintenance personnel. Many valves appear to have been

accepted as-is or accepted with minor lubrication corrective

action. No verification and documentation of individual valve

acceptability appears to have been performed. A majority of

the valves appear to have been accepted with no QC inspection.

(3) AP 650 revision 5, PM Program, categorized manual valves in QA

Class 1 systems that are required for controlld safe-shutdown

of the plant as Category 1 and Category 2A valves. Corrective

maintenance on Category 1 valves had been performed but the

licensee's program did not include a requirement or plan to

perform corrective maintenance on all Category 2A valves

before restart.

During this inspection, the inspector discussed the above noted

concerns with the Maintenance Manager and performed a sampling

inspection of the manual valves. The discussions and inspection

resulted in the following:

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(1) The Maintenance Manager provided the inspector with a copy of a ]

memorandum from the Nuclear Operations Manager to the l

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Maintenance Manager, dated May 4, 1987, that formalizes the

review and listing of 142 manual valves that are considered

critical for plant operations. The inspector had no further 2

concerns regarding this item.  !

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(2) The Maintenance Manager stated that inspections and corrective

maintenance had been completed on the 142 manual valves that

were critical for plant operations. Current procedures allow

for the QA Department to require QC verification of maintenance

activities regarding the manual valves but QC inspection was

not performed on the 142 manual valves that had been determined

to be critical for plant operations and none was planned at the

time of the initial discussion. On that basis, the inspector

selected 7 of the 142 valves for a sampling visual inspection i

to determine the adequacy of the maintenance department l

baseline inspections and corrective maintenance. The sample I

was selected based on the Senior Reactor Operator's I

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determination of valves on the list (142 critical manual

valves) that could be cycled by an auxiliary operator without

disrupting plant conditions or operations in progress and the

inspector's determination of relative importance of the valves.

The inspector also requested a QC inspector to perform the

inspections with the NRC inspector to determine the adequacy of ,

normal QC inspections of manual valves. The following are the l

valves inspected and the discrepancies observed:

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NSW-001 - (Nuclear Service Cooling Water Surge Tank Discharge

Isolation Valve).

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- Operation of valve handwheel causes a valve position-

indication reversed from the valve position direction

i marking _in the handwheel. As a result,'the auxiliary

operator, the QC inspector and the Maintenance

Engineer could not determine the-actual valve

position. 'In addition, during potential emergency

conditions, the valve condition could be diagnosed as

a stuck valve and forceful operation of the valve, if.

performed, could result in valve damage.

NSW-003 - (Nuclear Service Cooling Water Pump A Suction.  ;

Isolation Valve).

- Valve position indicator had no visible "open" or '

" closed" markings.

- Body and bonnet fasteners did not have full thread

engagement (i.e. not flush with the nut). No  !

apparent technical evaluation for acceptability was

provided by the licensee during the' inspection.

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BWS-036 "B" Boric Acid Pump Discharge Isolation.

- Body-to-Bonnet Leak.

- Packing Leak.

BSW-030 "B" Boric Acid Pump Suction Isolation.

- Packing Leak.

NRW-009 "A" Nuclear Service Cooling Water Hx Vent.

- No discrepancies observed. -!

FWS-046 - Auxiliary Fd Pump P319 Suction Isolation.

- No discrepancies observed.-

FWS-052 - Auxiliary Fd Pump P319 Flow Orifice Isolation. _

- No discrepancies observed.

As a result of the above inspection, QC issued NCR-6688 for the i

NSW-003 discrepancies and NCR-6687 for the NWS-001 i

discrepancies. As immediate corrective action, the Maintenance

Manager and QC Supervisor initiated a 100% reinspection of the  :

142 critical valves for the conditions noted above. t

(3) During the initial discussions, the Maintenance Manager stated

that the intent of AP 650 was not to categorize valves required

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for controlled safe shutdown of the plant in both Category 1

and Category 2A. The Maintenance Manager stated that a

revision to AP 650 would change the Category 2A definition and

would consequently only have valves required for controlled

safe shutdown of the plant in Category 1.

Pending licensee completion of satisfactory inspection of manual

valves and revision of AP 650 to reflect the current licensee

intent, RV-MA-4 was left open.

No violations or deviations were identified.

4. Licensee Action on Previously Identified Inspector Items

A. (Closed) Unresolved Item 85-23-03, Adequacy of IST Program

Procedures

.This item involves six separate procedure discrepancies and

weaknesses (Items 3.a - f in inspection report 50-312/85-23)inthe

licensee's Surveillance Procedures for implementation of the ASME,

Section XI, Inservice Test Program.

Item 3.d of the above report identified the absence of pump suction

pressure and bearing temperature acceptance criteria in SP 213.01.

This item was previously reviewed and is being separately tracked as

Unresolved Item 50-312/84-41-31 and is considered closed for the

purposes of this report.

Item 3.e was the subject of Violation 50-312/85-23-02 (Failure to

implement calibration controls for instruments used in Surveillance

Procedures) and of Deviation 50-312/87-11-01 (Failure to implement

corrective action for Violation 85-23-02). The licensee's actions

regarding Item 3.e will be reviewed in conjunction with their

responses to the Violation and Deviation and are considered closed

with respect to this Unresolved Item.

Item 3.f identified inadequacies in relief valve testing and was

cited as an example of Violation 85-23-04. Therefore, the

licensee's actions pursuant to this item will be reviewed in light

of their response to the violation and are considered closed with

respect to this Unresolved Item.

Item 3.a involves Decay Heat System return header isolation valves

DHS-015 and -016. SP 203.06 previously required the performer to

" record proper flow path and design flow provided through the

valves..." while the data sheet (Enclosure 6.2) required only a sign

I off for those valves to verify that the valves were locked open. i

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Revision 11 to SP 203.06 has been issued which requires DHS in

operation at 3,000 gpm flow (per Operating Procedure AP-8),

verification of the valve (s) in the open position, and recording of

the flow rate through the valves. The valve configuration is such

that full system flow rate (per AP-8) should be seen through the

valves. The inspector noted that no specific acceptance criteria is

provided but, based on the system configuration for test and the

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combined requirements of tl'e two procedures, the test method is now

acceptable. This portion of Unresolved Item 85-23-03 is considered

closed.

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Item 3.b - Main Turbine Throttle Stop (Trip) Valves TV-1 through

TV-4 are fast acting valves controlled by the electro-hydraulic

control system to trip the turbine and act as main steam isolation '

valves. The facility is not equipped with in line main steam

isolation valves upstream of the turbine steam chest.

SP 213.03C, " Turbine Throttle Stop Valve Fail Safe Test," provides

for full stroke testing of the valves in accordance with the IST-

program requirements and SP 214.02, Valve Test Information. Prior

revisions of SP 214.02 provided valve closure time requirements of  ;

3.3 to 5.6 seconds which were in conflict with the NRC Safety J

Evaluation Report, Relief Request PV-14, dated September 25, 1984,

which stipulated measurement of valve closure times with a maximum

closure time " typical of a new or recently reconditioned valve

(typically one second) plus two seconds...." SP 213.03C,

revision 1, included stroke time requirements of -3 seconds

consistent with the SER.

Item 3.b further noted that SP 213.03C did not require recording of

actual stroke time, only a verification that the valve stroked in

less than 3 seconds.

During the current inspection, SP 214.03, revision 6, was found to

include correct valve stroke time limits for TV-1 through TV-4 (1 - 4

3 seconds). This portion of Item 3.b is considered closed.

Revision 1 of SP 213.03C was still in effect, unrevised from the

previous inspection. The procedure did not require that stroke time

data be recorded but required only & verification signature that the

" valve moves from open to closed in three seconds or less."

TS 4.2.2, Inservice Inspection, requires inservice testing of ASME  !

Code Classes 1, 2, and 3 components to be performed as closely as

design permits in accordance with Section XI of the ASME Boiler and l

Pressure Vessel Code and applicable addenda as required by 10 CFR 1

50.55.a(g) except where specific written relief has been granted by

the Commission.

The NRC SER dated September 25, 1984, provides for relief from

certain aspects of ASME Section XI for valves TV-1 through TV-4 but

requires slow full stroke testing quarterly (without stroke time

measurement) and fail safe (fast) full stroke testing each cold

shutdown including stroke time measurement.

ASME Section XI, Article IWV3410 requires that the stroke time of  ;

power operated valves be measured to the nearest second or 10% of

the maximum allowable stroke time, whichever is less, whenever the

valve is full stroke tested. The data is required to be trend

analyzed for increases in stroke time (50% for fast acting valves)

to detect aonormal operation and invoke increased test frequencies

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or corrective action. No exemption is authorized by the NRC SER  :

except for the quarterly " slow" test. {

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Procedure 213.03C, revision 1, is applicable to both the quarterly {

" slow" full stroke test and the " fast" or fail safe test at cold ,

shutdown and does not discriminate between the two test requirements j

in either test method or data acquired. As indicated above, only

verification of stroke time less than 3 seconds is recorded.

Failure to measure, record and analyze stroke time data for TV-1

through TV-4 constitutes an apparent violation of TS 4.2.2 and the '

requirements of the inservice test program above (50-312/87-16-01).

Item 3.c also involves program relief request for RCP Seal Injection j

line isolation valves SIM-019, -020, -021, and -022. These are stop '

check valves which act as reactor coolant pressure boundary valves i

and, in the reverse (check) direction, as containment isolation  !

valves.

NRC SER Relief Exemption PV-3 states that the valves must be full l

stroke tested at each cold shutdown. Inspection 50-312/85-23 found i

that SP 203.03, " Quarterly Makeup and Purification System SFAS Valve l

Inspection and Surveillance Test: required only that the valves be l

" inspected"; no provisions for full stroke testing could be j

identified. j

During the current inspection, review of SP 203.03, revision 10, j

found that the applicable portion of the procedure was still as

stated above. This constitutes a further example of Violation

50-312/87-16-01 above. l

The licensee noted that, although SP 203.03 appeared unresponsive to

the above requirements, SP 205.02, Local Leak Rate Testing,

coincidentally resulted in these valves being closed and reopened

(on a nominal biennial schedule) as part of the 10 CFR 50, Appendix

J, Type C, containment isolation valve tests. Based on that fact, ,

the valves normally indicated flow conditions, and NRC authorized

relief from reverse flow testing of the containment isolation check I

valve function, the licensee considered the safety implications of

the testing oversight to be minimal. )

With respect to the two examples of noncompliance above, the

licensee advised that a revised IST Program Plan was under

preparation for the next 120-month cycle per 10 CFR 50.55a, that the

above test requirements and implementing procedure revisions were

included in that effort, and that the program was scheduled for )

submittal to NRC during June 1987. The licensee appeared to

consider that interim action to address the findings of the'1985

inspection was unwarranted. l

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The inspector further noted that, in addition to other related IST

Program findings, the Augmented System Review and Test Program

Inspection, Unresolved Item 50-312/86-41-26, found that the current ,

'IST Program had apparently expired on April 17, 1985 (as stipulated a

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in the'NRC SER above) and that the licensee.was operating without an-

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approved program. The licensee was unable to provide documentation 1

, that an extension past the above date.had been granted by NRC.  !

B. . (Closed) Unresolved Item 86-18-01, Acceptability'of engineering

evaluation of natural frequency of modified valve SFV-22006 and

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effects on dynamic piping analysis -;

.During Inspection 50-312/86-18, the.above valve. developed a body-to- I

bonnet coolant leak. The licensee installed a temporary, external j

clamping ring on the above valve to support injection of a temporary 1

l- sealant (Work Request 96425, December 8-9, 1985). Based upon review

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of the licensee's actions and the Updated Safety Analysis Report

, (USAR), Section 4.3, the previous inspector was unable to determine

whether the modification reduced the valve's natural frequency below

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,. the 20 Hz. limit of USAR Section 4.3. The licensee's evaluation did

L lnot' include a' specific review of the effect of the additional weight

l. on natural frequency although it determined that other related

l. effects were negligible.

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As a result of the above finding, SMUD Nuclear Engineering performed I

further engineering eva.luation as documented in Memo DLG-86-619, . I

dated October 17, 1986, finding that the temporary modification had l

no effect on the natural frequency of the valve.

The inspector reviewed calculation Z-PLS-M4093, Letdown Cooler Inlet'

Piping Stress Summary, revision 3, Appendix A, dated June 16, 1986,

and revision 4, dated December 18, 1986. I

These calculations showed that the original frequency was about 40

Hz. and was primarily dependent upon the cantilevered mass of the

valve actuator and its moment arm. The clamping ring was installed

on the' valve body (the supported " base" of the valve / actuator

combined mass) and had no effect on the natural frequency. The  ;

calculations further showed all other dynamic analysis

considerations to be well within design stress limits.

Permanent repairs were completed on the subject valve'on February

13, 1986, and the temporary clamping ring removed. Based on the

above noted licensee evaluation and permanent correction of the

condition noted, the unresolved item was closed.

No violations or deviations were identified.

C. (Closed) Unresolved Item 86-21-07, Motor Operated Valve Torque

Switch Setpoints not Incorporated in Drawings

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The previous inspection found that Drawing E1012, Motor Operated

Valve Data, did not include required torque switch settings for

valves:

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HV-20003, DHS ECCS Long Term Cooling Valve (DHR Test to-RB

Sump)

HV-23801, DHS ECCS Long Term Cooling Valve (DHR Test to RB

Sump)

HV-23802, Alternate Pressurizer; Spray Supply Valve

LV-36005, Auxiliary Boiler Isolation

LV-36505, Auxiliary Boiler Isolation

The inspector reviewed the following documents and discussed the

licensee's programs with licensee Maintenance, QC, Licensing and

contractor representatives:

QCI Tracking Item Recommendation Report 20.0136, April 27,

1987;

Memo, DLG-86-712, Nuclear Engineering to Nuclear Licensing,

November 4, 1986, forwarding Engineering Change Notices (ECNs)

for the subject valves;

ECNs R-0968C, R-0976AU, R-0976AV, R-0968AM, and R-0968AN for

the subject valves;

Coordinated Commitment ~ Tracking System Transmittal, CCTS #

T860818010C, Unresolved Item 86-21-07;

E-1012, Sheets 133, 166, and 167 incorporating Drawing Change

Notices 0 and 0A for HV-20003, -23801, -23802.

Procedure E-117C, Limitorque Valve Actuator SMB and SB Models

Electrical Refurbishment and Corrective Maintenance, revision

1.

The licensee has extensively revised Drawing E-1012. The drawing

formerly had multiple valves' data on each drawing sheet; the new

drawing format provides one sheet for each valve with considerably

more information than previously provided. l

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The current process for developing acceptable torque switch settings

involves:

Calculation of the maximum design basis differential pressure i

for each valve, then calculation of the maximum thrust and I

torque values. I

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Incorporation of data responsive to IE Bulletin 85-03, " Motor j

Operated Valve Common Mode Failures During Plant Transients Due l

to Improper Switch Settings," and IE Information Notice 86-29, i

" Effects of Changing Valve Motor Operator Switch Settings." l

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Revision of E-1012 sheets to delete old information and create

the new sheets for each valve with the newly calculated (and

other) data.

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Revision of affected elementary and wiring diagrams.

Performance of refurbishing and resetting per E-117C and the l

M0 VATS test program to confirm the switch settings.  ;

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Engineered torque switch setpoint values are now provided for the

subject' valves and will be confirmed via the maintenance and testing

program discussed above. The licensee's program appears acceptable

in this regard; this item is closed.

No violations or deviations were identified.

5. Onsite Follow-up of Written Reports of Non-routine Events at Power

Reactors

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A. (Closed) LER 80-46-T1, Overspeed Trip Lever on Auxiliary Feedwater

Pump Turbine Found Tripped

On November 4, 1980, the overspeed trip lever for Auxiliary

Feedwater Pump P318 turbine was found tripped during routine

surveillance. It was thought that the lever may have tripped due to

vibration from running the pump with its motor. drive. A daily

visual check of the trip lever was instituted.

On November 21, 1980, the trip lever was again found in the trip

position even though the pump had not been operated since the

November 4 instance; no cause could be determined.

Although the motor drive of the dual drive pump remained available,

the motor drive does not receive an automatic start signal and must

be started manually should an Auxiliary Feedwater initiation signal ,

occur with the turbine drive tripped.

LER 80-46 revision 1, dated February 9, 1987, described a visual

indication device installed on the turbine trip lever and

incorporation of the overspeed trip device status into the IDADS ]

computer monitoring system.

During this inspection, the IDADS computer point display Z1622 was

demonstrated by a licensee shift supervisor. The computer point is

annunciated in the control room and the status of the trip device

(tripped /untripped, alarm in/out, circuit signal quality, etc.) is q

video displayed and the point is equipped with both audible and '

visual alarms.

The turbine trip valve, pump and shift logs were inspected and the

use of the visual indicator was discussed with a Shift Supervisor.

The licensee currently includes a visual check of the indicator on

the shift logs but the LER advised that the licensee no longer

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considers this check to be~a mandatory requirement based on the

' addition of the IDADS point.

TheLinspector noted that the IDADS. system is considered non-safety

related and plant operations may continue if the' system is

unavailable or otherwise removed from-service. Casualty Procedure

C.39, Loss of IDADS, Revision 1, effective April 8, 1986, Enclosure

3.4, Item 3, provides for hourly verification.of the overspeed trip

reset condition during.IDADS outages. iThe root cause_for the AFW

pump trips could not be determined. .The" licensee's corrective.

action regarding monitoring of trip valve status appeared to be

appropriate for the condition reported. The LER was closed.

B. (Closed) LER 85-12-LO/L1, Loss of Reactor Building Integrity

i' While in hot shutdown conditions and during zero power physics

testing on June 16,- 1985, ' reactor building integrity was lost when l

both personnel air lock hatches were simultaneously opened.

' Containment integrity in these conditions is required by Technical

. Specification 3.6.1. The initial NRC followup of this report is

documented in'NRC Inspection Report 50-312/85-27 and identifies

NRC's W ncerns regarding. repetitive.similar occurrences (this was

the' fourth-such instance; reference LERs 83-19, 82-01, 81-13).

Reporting Timeliness

The licensee's initial written report committed to submit the.

results of a Root Cause Analysis for the incident to NRC by

Octobe r . 31^, .1986. The Root Cause Analysis, discussed below, was

completed on September 26, 1985, was approved on November 8, 1985,

and was issued via. memo GAC 85-895, on November 18, 1985. Its

results were not submitted to NRC until revision 1 to LER 85-12 was

issued on November 6, 1986.

The inspector further noted that the LER states that the event

occurred during " cold shutdown" while performing "zero power physics

testing." The plant was actually in " hot shutdown conditions" and-

during zero power physics testing.

The licensee's failure to meet a commitment to NRC is similar in

nature to: Deviation Item 87-11-01, failure to meet commitments for.

procedure revisions, issued on April 20, 1987. The licensee's

response to the Deviation item was pending at the time of this

inspection. The NRC letter and report forwarding the Deviation j

required the licensee to address general weaknesses in the '

licensee's program for tracking commitment schedules and notifying

NRC of any changes.

Accordingly, no enforcement item will be issued for the late

submittal of LER 85-12 followup'information on the basis that the

response to Item 87-11-01 above should also'be responsive to this

item. The SMUD licensing representative was advised of the above

and the need.to ensure that~the Item 87-11-01 response effectively

dealt with'all issues.

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Corrective Action Efficacy

As noted above, the subject event was the fourth in a series of

similar e_ vents. The licensee Root Cause Analysis, No. 85-14, .

identified the following root causes:. (1) inadequate maintenance, I

(2)' inadequate personnel training in door operation, and (3) prior

dismantling of the door audible alarm circuit. The report.further

identified multiple deficiencies and weakness contributing to the  !

-incidents and specified a Corrective Action Plan, including:  ;

(1) Implementation of training on proper operation of the door for

all vital badged personnel by December 31, 1985.

(2) -Installation of door rebound devices (snubbers) by the Cycle 9

Outage (next refueling).

(3) Reactivation of the audible door alarm circuit (no commitment

date provided).

(4) Establishment of maintenance requirements and intervals versus

door usage (no commitment date provided).

(5) Conduct'a " human factors" review of door handwheel orientation, j

use of signs and labels, etc. (no commitment date provided).  !

(6) Evaluate the desirability of performing modifications

recommended by the door manufacturer (no commitment date

, provided). j

Revision 1 to LER 85-12 discussed Items 1 through 5 above stating

that they would be accomplished by the Cycle 8 refueling outage,  !

currently scheduled for September 15, 1988. Item 6 was not

addressed. The licensee report further indicated that routine

testing of the door and airlock per the Technical Specification (TS) l

Surveillance Requirements will " continue to provide assurance of the

containment integrity of the reactor building personnel access

door." The LER notes that TS 3.6.1 and SP 205.02 require that the

doors be tested only within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each opening.

-The inspector further noted that the Plant Buildings and Structures

System Investigation Report (SIR), revision 0, performed per QCI-12,

Systems Review and Test Program (SRTP) identified this matter as

Problem No. 5 (Tracking No. 20.0430). The SIR identified only items

1, 3, and 5 of the Root Cause Analysis as required resolutions. The

SIR noted that, "...until all equipment and programmatic changes are

fully implemented, specially trained personnel must perform airlock

door manipulations during those outage conditions that require i

containment integrity...." The SIR categorized the item as Priority i

3, not requiring further action prior to restart. The inspector

could identify no such actions planned or in progress.

In an attempt to reconcile the differences between the LER, the SIR

and the Root Cause Analysis Corrective Action Plan and to establish

whether any further interim actions were planned, the inspector 1

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requested additional information from the SMUD licensing

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representatives and the. Plant Buildings and Structures System

Engineer. Additional information was not available from the plant

staff and a meeting was requested with the Plant Manager to discuss

the-inspector's concerns.

During.a meeting with the Plant Manager and his staff on May 7, the

inspectors:were advised that the containment airlock vendor will be .

contracted to provide door refurbishment, assistance in establishing j

maintenance criteria, and maintenance personnel training. These l

activities are scheduled for the summer, 1987, prior to plant

restart. .

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The vendor apparently believes that door refurbishment and

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maintenance will essentially eliminate the mechanical problems which I

contributed to this and the prior incidents. The licensee advised

that following these actions, door operation would be evaluated and i

additional actions taken, as necessary, if the problem indicators

recurred. The Plant Manager emphasized that the Systems Review and

Test Program had established that the airlock problems had been-

categorized as Priority 3 actions and that the above interim actions i

were not to be considered as a commitment to implement any actions

prior to plant restart.

The inspectors requested that the above interim actions be

documented as formal commitments in a revised, supplemental LER.

The inspectors stipulated that the prior door failure scenarios had

the potential for simultaneous inner and outer airlock door failures

which could prevent either or both doors from being quickly

reclosed, e.g.', . failure of actuating gears, interlocks, etc. , as

previously experienced. This would result in a serious, extended

breach of containment integrity. l

The Plant Manager stated that the SRTP disagrees with the

inspector's conclusions, that such a commitment would be an-

exception to licensee policy subject to approval by the Chief

Executive Officer (CEO), Nuclear, and that he would advise the l

inspectors of the CE0's position.

The inspectors advised the Plant Manager that the licensee's

proposed long-term corrective actions will not be implemented for

one complete operating cycle and no interim or compensatory actions

are planned and the licensee's plans are considered inadequate to

prevent recurrence in a timely manner. Routine surveillance testing

stipulated by the LER is considered. inadequate to prevent recurrence

and would serve only to provide after.the fact identification of

improperly sealed doors with the alarm (and in some past cases,

interlocks) ineffective.

Subsequent to the NRC exit meeting, on May 12, 1987, the Plant

Manager informed the inspector by telephone that LER 85-12 will be

revised to provide the following corrective action prior to restart:

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The containment airlock vendor will be contracted to train

licensee personnel on vendor experience with the doors.

A " maintenance tune-up" and preventative maintenance will be

performed to provide recommended corrective action for the

doors.

Recommended corrective actions will be evaluated for required

performance prior to restart.

LER 85-12 revision 0 will be administratively closed. Revision 1 of

LER 85-12 will be administratively closed upon receipt of revision 2

of LER 85-12. Revision 2 of LER 85-12 will be evaluated for further

inspection followup.

No violations or deviations were identified.

C. Licensee Event Reports 85-16, 86-10, 87-13, 87-76 - Cable Raceway

Tracking System (CRTS) Related Items

On April 3,1987, the licensee submitted to the NRC a Wire and Cable

Program Description and Action Plan. On May 1, 1987, RV provided

the licensee written comments regarding the plan. The inspector met

with the licensee to discuss the RV comments. In addition, on May

6, 1987, the licensee met with representatives from NRR and RV to I

discuss the program. As a result of the RV comments and comments

during the May 6, 1987, meeting, the Engineering Department Manager

stated that a revision to the CRTS Program would be submitted to the

NRC.

No violations or deviations were identified.

6. 10 CFR Part 21 Reports ,

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A. (Closed) 10 CFR 21 Report 86-03, Incorrectly assembled Woodward

Governors supplied by Transamerica Delaval, Inc. (TDI) for Emergency

Diesel Generators (EDGs)

In a letter dated August 10, 1983, TDI advised that Woodward

governor internals provided on new TDI EDGs at Rancho Seco had been

j incorrectly assembled. If the governors were not~ modified, the

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possibility existed that the engines may overspeed on an automatic

start.  ;

The inspector reviewed the following documents confirming that the

governors had been returned to the manufacturer, properly rebuilt

and reinstalled on the EDGs.

TDI Trip Report, Visit to Rancho Seco, September 16, 1983,

identifying the initial governor problem and recommending

rework by Woodward.

SMUD letter to TDI, September 23, 1983, confirming TDI

arrangements for reworking the governor and requesting TDI's

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approval of a SMUD governor removal / replacement procedure

(enclosed).

TDI letter to SMUD, October 28, 1983, confirming governor

rework arrangements and approving the above SMUD procedure.

SMUD Shipping Notice No. 50557/P.O. BA1123, December 15, 1983,

documenting shipment of governors to Woodward Governor Co. for

rework.

SMUD Expeditors Report, March 19, 1984, documenting shipping

plans for return of governors.

Receipt Inspection Plan No. QA1203-81, Diesel Generator Parts,

March 2, 1984, documenting receipt of Woodward governors.

TDI Certificate of Compliance for gov'ernors, March 30, 1984.

Receiving Inspection Data Report No. QA1203-81, May 21, 1984,

documenting results of receipt inspection.

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Quality Verification Document Requirements checklist, March 30, l

1984, documenting verification of test reports, seismic l

qualification documentation, certificate of conformance, etc.

The above documentation confirmed that the governors had been

repaired as required to correct.the problem identified by the

subject 10 CFR 21 Report. The inspector verified, by visual

inspection of serial numbers and the installed governor units that

the governors are installed on the engines. This item is closed.

No violations or deviations were identified.

B. (Closed) 10CFR21 Report 86-13-P, Missing Lockwelds on Anchor

Darling Check Valves.

The Anchor Darling Valve Company reported that lock welds were found

missing on 150 and 300 class check valves' hinge support, hinge

support capscrews, and hinge support / bonnet interface. The report

recommended inspection and repair techniques.

This item was previously reviewed during Region V inspection

50-312/86-38. That inspection confirmed that the licensee had

implemented an inspection and repair program for twenty-two

potentially affected valves. Six valves had been inspected at that j

time.

By the time of this inspection, the licensee had inspected fifteen

(total) valves. Six of the fifteen were found to have either

cracked tack welds or minor cracks in hinge bushings, all of which

have been repaired. The remaining nine valves inspected were

satisfactory.

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The licensee's' evaluation of the as-found conditions determined that {

none of the defects would have affected valve. operation. The cracks '

were characterized as small and in locations which would not result

in loss of valve assembly integrity. The inspector reviewed these

determinations with the responsible licensee mechanical maintenance

engineer and found the licensee's findings and conclusions

reasonable. j

Inspection of the remaining valves is being tracked by the  !

licensee's Coordinated Commitment Tracking System (No..T8611061010) l

and is scheduled for completion prior to restart. Maintenance Work  !

Requests have been issued for all the valves. This item is closed.  ;

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No violations or deviations were identified.

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C. (0 pen) 10 CFR 21 Report 86-14-P, Failure of Automatic Sprinkler

Corp. of America (ASC0A) Model C Deluge Valves

This report involves the potential for failure of Model C fire

protection deluge valves to actuate on demand due to binding of

bronze valve latch arms. The latch arm problem applies only to six

inch valves. ASC0A notified NRC and potentially affected licensees

via the various letters below:

June 4, 1984, Letter to NRC, Additional Information re IE

Information Notice 84-16, " Failure of Automatic Sprinkler

System Valves to Operate."

November 19, 1985, Letter to NRC, Model C Deluge Valves.

December 1, 1985, Letter to licensees, Model C Deluge Valves.

December 12, 1985, Letter to licensees, Model C Deluge Valves.

April 7, 1986, 10 CFR 21 Report to NRC, Model C Deluge Valves.

April 18, 1986, Letter to NRC, Followup on April 7, 1986 10 CFR

21 Report.

• November 5, 1986, Letter to licensees, Model C. Deluge Valves .

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and Mercury Check Devices.

(* Note: In this letter, ASC0A also advised licensees of

p;tential failure of Mercury Check Devices due to plastic parts

leaking. The Mercury Check Devices are used to sense

temperature rate-of-rise to prevent spurious sprinkler syster.,

actuation on slow ambient temperature increases not resulting

from fire.)

The licensee has received and evaluated the above correspondence. '

The licensee's fire insurance underwriter has also addressed

recommendations and licensee actions for the above in an August 1986

Loss Prevention Report. The licensee has determined that thirteen

Type C valves are installed at Rancho Seco. Six of the valves are

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the 6" size subject'to the potential latch failures. The remaining 1

seven valves are 2-1/2" valves which ASC0A advises are not affected {

by the latch failures but may be. subject to the Mercury Check Device )

problems.

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The. licensee has determined by review of procurement records'that d

none'of the subject valves (by serial number) at Rancho Seco were  !

initially purchased with bronze latch arms. The licensee is further

performing a review of valve maintenance history to identify past

latch failures and whether bronze _ replacement parts were ever-

installed. To date, about 75% of the maintenance records have'been

. reviewed with no latch failures nor bronze part installations found.

-The licensee plans to physically inspect the valves (see further

discussion below) to further confirm the absence of bronze latch arm

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parts. l

The various ASC0A letters recommend that the valve latch arms be ,

lubricated periodically. The licensee issued. Work Request 111958 on

March 19, 1986 to accomplish this. As.of May 1, 1987, this

lubrication'had not yet been accomplished; Work Requests for

inspection lof.6" valve latch arms and replacement of Mercury Check 1

Devices are planned but not' issued-and are to be performed l

concurrently with the lubrication.

The cognizant' licensee fire protection engineer advised that these

items are not prioritized as plant restart prerequisites and will be j

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accomplished by September 1987. This is consistent with the Fire '

Protection System Investigation Report prioritization of the item.

This 10 CFR 21 report will remain open pending further evidence of

implementation of the licensee's program lubrication, inspection,

and check device replacement.

.No violations or deviations were' identified.

D. (Closed)'10 CFR 21 Report 86-15-P, Damaged Motor Leads in Limitorque

Valve Operators

(Closed) 10 CFR 21 Report 86-18-P, Cracked Limit Switch Rotors in

Limitorque Valve Operators

(Closed) 10 CFR 21 Report 87-07-P, Warped Limit Switch Rotors in

Limitorque Valve Operators

These reports detailed defects in Limitorque motorized valve-

operators. Report 86-15 involved damaged leads found'in replacement

valve operators. Reports 86-18 and 87-07 identified defects in

limits switches during inspections of installed valves with various

service times.  ?#

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The licensee has incorporated the above report items into the

ongoing motor operated valve refurbishment program. Reports 86-18  :

and 87-07 were identified by this licensee,

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Provisions to address each of the above are included in Procedure

E-117C, Limitorque Valve Actuator SMB and SB Models Electrical

Refurbishment and Corrective Maintenance Procedure, revision 1.

Instructions for inspection and replacement of damaged motor and

brake leads are included in Sections 6.2.7.8 and 6.2.9.

Instructions for inspection and replacement of cracked or warped

. limit switch rotors (and unqualified items) are included in Section

6.2.1.1.13.1.

The inspector reviewed completed work packages for the following

valves, confirming that: the applicable procedure sections were

accomplished, deficiencies were identified and corrected, proper

replacement parts were used and properly installed, and the

activities were subject to Quality Control inspection. The licensee

advised that, to date, 27 limit switch rotors have been replaced

(count based on stores withdrawal records). j

Work Request 121773, Valve HV 26046, DHS Cross Tie "A",

completed April 24, 1987

Work Request 120371, Valve HV 26007, "A" DHS Cooler Outlet to

Pump, completed April 24, 1987

Work Request 120364, Valve SFV 50009, NSW from A500A, completed

April 24, 1987

Work Request 120355, Valve HV 26105, Emergency Sump to DH Pump

261A, completed April 24, 1987.

No discrepancies were identified. The licensee has acceptably

incorporated the considerations of the above reports into their

programs. The above noted 10 CFR 21 items are closed.

No violations or deviations were identified.

E. (Closed) 10 CFR 21 Report 86-25-P, Unqualified Replacement Terminal

Strips for Limitorque Motor Actuators

The subject report (filed by Arkansas Power and Light) advised that

Buchanan 724 terminal strips had been supplied by Limitorque as in

kind, qualified replacements for Buchanan 524 strips installed in

valve motor operators. Environmental qualification could not be

established for the Model 724 strips.

The inspector reviewed SMUD Memo BK 86-089, dated December 22, 1986,

which documented the environmental qualification status of this

facility's motor operated valves. The memo noted that all terminal

blocks had been removed from 70 of 99 of the valves and the affected

leads spliced with environmentally qualified splices. Twenty-five

additional valves had not been previously modified and are being

modified per ECN R-0699 during the current outage.

The inspector also reviewed Calculation Sheet No. L-EQP-E0133, dated

September 25, 1986, which identified the modifications for the

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subsequent modifications and included replacement of the internal

wiring and terminal blocks with qualified wiring and splices. Work

packages for selected valves were reviewed as documented elsewhere

in this report, confirming that the valve refurbishment procedures,

as implemented resulted in splice installation. Additionally,

Procedure NEP 4118 includes provisions to ensure that newly l

purchased valves are site modified to meet the above configuration '

criteria prior to service. Accordingly, this 10 CFR 21 report is

not applicable to Rancho Seco. ,

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No violations or deviations were identified. l

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7. Task: Allegation ATS No. RV-87-A-024 (Closed) ]

A. Characterization  !

A Rancho Seco employee was working on radiation monitoring systems

without previous experience or training. ,

B. Implied Significance to Design, Construction or Operation

Work on safety-related systems was being performed by unqualified

personnel.

C. Assessment of Safety Significance

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A review of SMUD records confirmed that the alleged individual was

an employee of Volt Technical Services performing contractor work

for the Rancho Seco Nuclear Maintenance Department in the Instrument

and Control (I&C) area. Training and Qualification records provided

by SMUD and interviews with the individual and two levels of SMUD

supervision indicate that-the individual had been trained and

qualified at another_ Region V Reactor Plant facility and was

undergoing an on-the-job training program for Rancho Seco I&C

technicians. The individual and two levels of his supervision

stated that he had not performed work on the radiation monitoring j

system. In addition, statement from the individual and his

supervisors indicate that all safety related work by contract

personnel are assigned with accompanying SMUD personnel.

D. Staff Position

Based on the records provided by the licensee and interviews with

the individual and two SMUD supervisors, the allegation was not i

confirmed, and is closed.

E. Action Required l

None. 1

8. Exit Interview I

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The inspection scope and findings were summarized on May 8, 1987, with l

those persons indicated in paragraph 1 above. The inspector described '

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'the areas inspected and discussed in detail the inspection findings. No

dissenting comments were received from the licensee. The following new

item was: identified during this inspection:

Violation 87-16-01 - Inadequate Inservice Testing Procedures for

Valve Full Stroke Testing, paragraph 4.A,

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