ML20202G278

Insp Rept 50-271/98-14 on 981122-990104.No Violations Noted. Major Areas Inspected:Operations,Maint & Engineering

ML20202G278
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	01/28/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20202G265	List: ML20202G261 ML20202G278
References
50-271-98-14, NUDOCS 9902050078
Download: ML20202G278 (31)

See also: IR 05000271/1998014

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION I

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Docket No.

50-271

Licensee No.

DPR-28

Report No.

98-14

. Licensee:

Vermont Yankee Nuclear Power Corporation

Facility:

Vermont Yankee Nuclear Power Station

. Location:

Vernon, Vermont

Dates:

November 22,1998 - January 4,1999

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Inspectors:

Brian J. McDermott, Senior Resident inspector

Edward C. Knutson, Resident inspector

William A. Maier, Emergency Preparedness Specialist

Richard P. Croteau, Project Manager, NR3

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Douglas A. Dempsey, Reactor Engineer

Thomas G. Scarbrough, Senior Mechanical Engineer, NRR

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Approved by:

Clifford J. Anderson, Chief, Projects Branch 5

Division of Reactor Projects

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2050078 990128

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ADOCK 05000271

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EXECUTIVE SUMMARY

Vermont Yankee Nuclear Power Station

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NRC Inspection Report 50-271/98-14

This inspection included aspects of licensee operations, engineering, maintenance, and plant

support. The report covers a six week period of routine resident inspector activities, and

includes the results of in-office procedure review by an emergency preparedness specialist. This

report also provides the results of a motor-operated valve inspection conductad the week of

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November 16,1998.

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Operations

VY was slow to pursue corrective action after the NRC identified a degraded high voltage

power supply with the potential to affected the operability of a TS-required instrument in

the reactor building ventilation isolation system. Once initiated, VY's corrective action was

prompt and adequately resolved the degraded condition. (Section 01.1)

VY failed to recognize that a long standing practice of allowing manual containment

isolation valves to be opened under administrative controls was in conflict with the

Technical Specifications. A November 1998 procedure change review was weak be ;ause

it invoked this practice for draining the torus and was a missed opportunity to identify the

problem. VY's practice did not compromise plant safety and the licensee promptly

submitted a TS change to correct the problem. (Section O3.1)

Maintenance

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The maintenance activities observed during this period were performed well. Workers

demonstrated appropriate radiological control and foreign material exclusion control

techniques. Good supervisory oversight, system engineering involvement, and

radiological protection support were observed. (Section M1.1)

The surveillance activities observed during the period were correctly performed. However,

in one case the multiple procedures which control the standby gas treatment system

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charcoal sample removal had the potential to cause errors. Activities were well controlled

and coordinated by the control room operators. (Section M1.2)

Maintenance personnel initiated work on the wrong standby gas treatment system filter

train and caused the entire system to be declared inoperable for a short period of time.

The error was identified by the licensee and appropriate corrective actions were initiated,

including a Maintenance department work stand down. The workers' failure to follow the

maintenance procedure is a violation of TS 6.5 and this issue was treated as a non-cited

violation. (Section M1.3)

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Executive Summary (cont'd)

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W identified that a pressure switch for the HPCI steam supply isolation logic had been

isolated during corrective maintenance and had not been properly returned to sewice.

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Because the switch had been depressurized, the low steam line pressure isolation would

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have functioned, if required. The failure to follow maintenance procedures was

determined to be a Non-cited Violation based on an assessment of the safety significance

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of the condition and W's corrective actions. (Section M1.4)

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The "C" residual heat removal service water pump failed inservice test acceptance criteria

for differential pressure on two occasions. Although immediate corrective actions restored

acceptable performance, W developed an operability justification to address the

degradation that was obsewed. The operability justification was adequate and W

management placed priority on the resolution of this degradad condition. (Section M2.1)

Primary containment isolation valve HPCI-16 failed to stroke closed during an inservice

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test due to the failure of the torque switch in its motor actuator. Appropriate immediate

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actions were taken in response to the test failure, a good evaluation was made to assess

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the generic implications of the problem and the failed torque switch was replaced.

Although the failure could have prevented full seating of the valve, the valve would have

closed enough to mitigate a high energy line break event. An inspector follow-up item was

initiated to track NRC review of W's final disposition of this issue. (Section M2.2)

W's approach to the Maintenance Rule requirements for assessing the effects of out-of-

service equipment on overall safety functions is consistent with NRC-accepted guidance.

However, implementing procedures lacked positive confirmation that alternatives to the

pre-analyzed work had been evaluated in accordance with the program expectations.

(Section M3.1)

W's methods for acquiring Maintenance Rule performance monitoring data are generally

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effective. However, the recording of unplanned equipment outages and the screening of

Maintenance Rule-related Event Reports are two areas where the accurate collection of

data may be challenged. (Section M3.1)

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Enaineerina

Positive aspects of the Generic Letter 96-05 periodic verification program for motor-

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operated valves (MOVs) were observed, including: (1) development of more efficient test

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techniques, (2) implementation of a motor test program, and (3) an aggressive motor-

actuator lubrication and refurbishment schedule. However, several aspects of the periodic

verification program, such as program documentation and MOV performance degradation

rates were as yet undeveloped. (Section E1.1)

Design-basis thrust calculations for two MOVs were not revised to reflect dynamic test

information, resu! ting in the calculations not reflecting the actual plant configuration. This

condition was an example of poor configuration management. (Section E1.1)

The inservice test failures of two scram discharge volume drain valves led to the

identification of problems with the new valve / actuator design installed during the 1998

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- Executive Summary (cont'd)

refueling outage. Errors were identified in the safety evaluation and there is a lack of

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design information from the vendor. In accordance with NRC guidance, this issue, which

may represent a violation of NRC requirements will remain open for a reasonable time to

allow the licensea to develop its corrective actions. (Section E2.1)

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TABLE OF CONTENTS

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EXECUTIVE SU MMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

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TABLE O F CONTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v

- Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1. Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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Conduct of 0pe rations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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O1.1 Area Radiation Monitor Power Supply Failure . . . . . . . . . . . . . . . . . . . 1

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Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . , . 2

03.1 Administrative Control of Manual Containment Isolation Valves . . . . . . 2

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Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

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08.1 In-Office Review of LERs Related to Operations . . . . . . . . . . . . . . . . . 3

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l l . Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

M1

' Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

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M1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

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M1.2 Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

M1.3 Standby Gas Treatment System Maintenance . . . . . . . . . . . . . . . . . . . 6

M1.4. HPCI Low Steam Pressure Isolation Switch Not Returned To Service . 7

M2.

Maintenance and Material Condition of Facilities and Equipment . . . . . . . . . . . 9

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M2.1 ' Residual Heat Removal Service Water Pump "C" Low Differential

Pressu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

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M2.2 Motor-Operated Valve Torque Switch Failure . . . . . . . . . . . . . . . . . . . 10

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M3

Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 11

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M3.1 Maintenance Rule implementation Review . . . . . . . . . . . . . . . . . . . . . 11

M8

Miscellaneous Maintenance lssues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

M8.1 In-Office Review of LERs Related to Maintenance . . . . . . . . . . . . . . . 14

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Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

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E1.1

Implementation of Generic Letter (GL) 96-05, " Periodic Verification of

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Design-Basis Capability of Safety-Related Motor-Operated Valves" . . 15

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Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 20

E2.1

Scram Discharge Volume Drain Valve Failures . . . . . . . . . . . . . . . . . 20

E8

Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

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E8.1

Review of Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

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E8.2 Review of VY Cycle 19 Operating Report . . . . . . . . . . . . . . . . . . . . . . 22

IV. Plant Support . . . . . . . . . . . . . . . . . . . . .................................... 23

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P3

EP Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

P3.1

In-Office Review of Emergency Plan Implementing Procedes . . . . . 23

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Table of Contents (cont'd)-

V. Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

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Exit Meeting Sum mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

ATTACHMENTS

Attachment 1 - List of Acronyms Used

Attachment 2 - ltems Opened, Closed, or Discussed

Atthchment 3 - Emergency Response Plan and Implementing Procedures Reviewed

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Report Details

Summarv of Plant Status

Throughout the inspection period, Vermont Yankee (VY) was operating at 100 percent power,

with few exceptions. Minor power reductions were made in support of surveillance testing and a

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l. Operations

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. Conduct of Operations'

01.1 Area Radiation Monitor Power Suoolv Failure

a.

Inspection Scooe (71707)

' Area radiation monitor (ARMS) for the reactor building ventilation and refuel floor provide

input to the reactor building ventilation isolation. The detectors are powered by a safety

class high voltage power supply, ES-17-451B. During a routine control room tour, the

inspector noted that the output voltage of ES-17-451B was significantly lower than

expected and informed the licensee. The inspector assessed VY's response to the

degraded condition.

b.

Observations and Findinos

- On November 23, the inspector observed that the voltage meter on high voltage supply

ES-17-451B indicated 440 VDC, whereas the normal voltage for such a power supply was

approximately 600 VDC. The inspector related this observation to an on-shift licensed

operator, and again to the Operations Planning Group.

On November 24, the inspector observed that the power supply voltage had decreased to

420 VDC, confirming that the unit was degrading. Although a work order request had

been initiated the previous day, the voltage indication had not yet been verified and the

potential effect of low voltage on the operability of the associated ARMS had not been

thoroughly evaluated. After the inspector questioned the equipment operability, system

' engineering determined that the affected ARMS should not be considered operable

because the lower detector voltage could have a negative impact on detector sensitivity.

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The two ARMS were declared inoperable at 7:30 a.m. on November 24. Technical Specification 3.2, " Protective Instrument Systems," Table 3.2.3, requires that the reactor

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building ventilation system be isolated and the standby gas treatment system operated if

the subject ARMS are not available for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In response to this problem, VY performed as-found calibration checks on the affected

ARMS and determined that they were operating satisfactorily. ES-17-451B was replaced

later that day, and the ARMS were declared operable at 4:00 p.m.

' Topical headings such as O1, M8, etc., are used in accordance with the NRC standardized reactor

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inspection report outline. Individual reports are not expected to address all outline topics.

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Since the as-found calibrations of the affected ARMS were satisfactory, a condition

requiring entry into the action statement never actually existed; therefore, no violation of

TS occurred. Nonetheless, VY's actions in response to the identification of this problem

were slow. NRC inspection manual part 9900, " Operable / Operability: Ensuring the

Functional Capability of a System or Component," states that the timeliness of operability

determinations should be commensurate with the safety significance of the issue, and that

the allowed outage times ' contained in TS generally provide reasonable guidelines for

safety significance. In this case, VY did not perform a thorough investigation of the

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condition or evaluate the potential consequences of the condition for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

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after the problem was identified.

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Conclusions

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VY was slow to pursue corrective action after the NRC identified a degraded high voltage

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power supply with the potential to affected the operability of a TS-required instrument in

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the reactor building ventilation isolation system. Once initiated, VY's corrective action was

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prompt and adequately resolved the degraded condition.

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Operations Procedures and Documentation

O3.1 Administrative Control of Manual Containment Isolation Valves

a.

Inspection Scooe (71707)

On November 23,1998, the inspectors noted that VY revised procedure OP-2123 to allow

torus water level control using manual valves, including a normally locked closed

containment isolation valve (CIV). The inspector reviewed this procedure against the

FSAR description and TS requirements for primary containment.

b.

Observations and Findinas

A partial revision of OP-2123, " Core Spray," was approved on November 20,1998, to

allowed the use of a draln line on the core spray (CS) suction pipe, and temporary hose, to

drain water from the suppression pool to an equipment drain sump. The drain line has two

manual valves in series and is normally capped. The procedure change allowed a local

operator to uncap the line, unlock and open the manual CIV, and open a second

downstream manual valve to establish the drain flow path. OP-2123 also requires the

auxiliary operator to remain at the valves and to have direct communication with the

control room. This activity was first performed on November 20,1998, with the

requirements for primary containment integrity in effect.

TS 3.7.2 requires primary containment to be maintained at all time when the reactor is

critical. TS 1.N defines Primary Containment Integrity and, in part, states, all manual

containment isolation valves on lines connecting to the containment which are not required

to be open during accident conditions are closed. The inspector discussed the apparent

conflict between the revision to OP-2123 and TS with VY management.

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VY management initiated ER 98-2092 to determine the cause of this event and to develop

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corrective actions. VY subsequently identified several other activities where manual CIVs

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are opencd under administrative controls, including TS required surveillance testing. As

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an immediate corrective action, Operations management required Operators to document

entry into the TS 3.7.2,24-hour action statement for loss of primary containment integrity

when manual CIVs were opened. On December 11,1998, VY submitted a TS change

request to allow administrative control of manual containment isolation valves and

eliminate the need to enter the TS action statement for loss of primary containment

integrity.

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The inspector noted that VY's practice of using administratie controls for these valves

was consistent with standard Technical Specifications and did not compromise plant

safety. However, VY's practice was in conflict with their TS and should have been

recognized. The inspector concluded that VY's failure to recognize the conflict between

the TS wording and the procedure revision as a weakness in the procedure review and

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approval.

VY procedure AP-0152 requires the preparation of the Control Room Shift Turnover

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Checklist, which includes a log (VYAPF 0152.02) for inoperable Technical Specifications

systems or components. Step 4.a.1. of this procedure requires the use of this log when a

system or component is inoperable. VY's past failure to document entry into the TS action

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statement during use of the manual CIVs is a violation of VY procedures. This failure

constitutes a violation of minor significance and is not subject to formal enforcement

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action.

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c.

Conclusions

VY failed to recognize that their long standing practice of allowing manual containment

isolation valves to be opened under administrative controls was in conflict with the

Technical Specifications. A November 1998 procedure change review was weak because

it invoked this practice for draining the torus and was a missed opportunity to identify the

problem. VY's practice did not compromise plant safety and the licensee promptly

submitted a TS change to correct the problem.

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Miscellaneous Operations issues

08.1

In-Office Review of LERs Related to Ooerations (90712)

An in-office review of LERs was performed to assess whether further NRC actions were

required. The adequacy of the overall event description, immediate actions taken, cause

determination, and corrective actions were considered during this review. The following

issues were closed-out based on the in-office review.

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(Closed) LER 98-020-01: Inadequate Equipment Control Practices Result in Two

Mispositioned isolation Valves Allowing Degradation of Primary Containment Integrity

Supplemental LER 98-020-01 provided VY's conclusion regarding the root cause of the

mispositioned valves and a description of the long term corrective actions. VY's

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investigation determined that inadequate equipment control practices led to the problem.

Corrective actions planned for the event included a review of independent verification

practices, briefings of Operations crews, and changes to improve the clarity of valve

restoration documentation. The licensee's corrective actions described in the

supplemental report appeared adequate. The event was evaluated in NRC Inspection

Report 50-271/98-10, the original LER tvas reviewed, and the problem was dispositioned

as a non-cited violation. The inspector concluded that no additional action is required and

this LER is closed.

(Closed) LER 98-019-00: Off-Normal System Alignment Following a Plant Trip Which

involved the Loss of a Reactor Water Recirculation System Pump and Reactor Water

Thermal Stratification Results in a Spurious Shutdown Cooling Isolation

This event occurred on June 10,1998, during transition to the residual heat removal

(RHR) system shutdown cooling (SDC) modo of operation. Initiation of RHR flow caused

a transient pressure increase that exceeded the high pressure isolation setpoint and

resulted in automatic closure of the SDC suction isolation valves. Similar events had

occurred several times in the past, and had been addressed through operating procedure

changes and operator practices to gradually initiate flow. The precise cause of this event

was still being investigated at the time that the LER was issued, with the results to be

submitted in a supplemental report; however, as indicated by the LER title, the off-normal

plant alignment leading up to the event and resultant thermal stratification are likely

contributors. No water hammer occurred due to this event, and actual reactor pressure

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did not exceed the automatic isolation setpoint. Therefore, this LER is closed.

11. Maintenance

M1

Conduct of Maintenance

M1.1 Maintenance Observations

a.

Insoection Scope (62707)

The inspector observed portions of plant maintenance activities to verify that the correct

parts and tools were utilized, the applicable industry code and technical specification

requirements were satisfied, adequate measures were in place to ensure personnel safety

and prevent damage to plant structures, systems, and components, and to ensure that

equipment operability was verified upon completion of post maintenance testing,

implementation of the Maintenance Rule Program was also reviewed when applicable to

these activities.

b.

Observations and Findinas

The inspector observed portions of the following activities and reviewed a sample of the

administrative controls for the maintenance.

Control rod drive pump "B" rotating assembly replacement per work order 97-

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11585, observed on December 8 and 12.

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-- Observed preliminary disassembly to support the pump casing lift the following day,

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Work was being pc rformed in a contaminated area with dedicated RP support and full

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time foreman supavision. No deficiencies were noted.

-- On December 12, observed preparations to land the pump casing. Level 3 FME

controls were in effect and the material accountability log was being properly maintained.

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No deficiencies were noted.

Valve CRD-33B repair, observed on December 12.

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-- Attended the pre-job brief for CRD-33B disassembly and noted that the appropriate

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personnel were in attendance. Emphasis was placed on the time that would be allotted for

the work (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, based on 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> between draining the scram discharge volume). In

addition, operations personnel were to have a separate brief prior to valve disassembly to

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discuss contingencies for draining during the maintenance.

-- Observed valve removal, disassembly, and reassembly. There were no obvious

problems with the valve internals. A vendor representative and engineering were at the

work site, and the job received full time RP coverage. No deficiencies were noted.

Safety related fuse replacements in the control room performed on December 18,

1998, under multiple work orders.

-- Observed the pre-evolution brief and noted a good emphasis was placed on confirming

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the accuracy of work documents and concurrent verification of the maintenance activities.

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-- System engineering support was present and actively involved in the oversight of the

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activities.

c.

Conclusions

The maintenance activities observed during this period were performed well. Workers

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demonstrated appropriate radiological control and foreign material exclusion control

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techniques. Good supervisory oversight, system engineering involvement, and

radiological protection support were observed.

M1.2 Surveillance Observations

a.

Inspection Scope (61726)

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The inspector observed portions of a surveillance test to verify proper calibration of test

instrumentation, use of approved procedures, performance of work by qualified personnel,

conformance to Limiting Conditions for Operations (LCOs), and correct post-test system

restoration.

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b.

Observations and Findinas

Portions of the following surveillance activities were observed:

"A" and "B" emergency diesel generator monthly surveillances, observed on

November 24-25. No deficiencies were noted.

High pressure coolant injection (HPCI) quarterly flow sunteillance, observed on

November 30.

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-- The HPCI steam supply isolation valve, V23-16, failed during this test. Refer to

section M2.2 of this report.

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"B" star'dby gas treatment system charcoal filter test cell change-out per OP-4501,

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" Filter Tosting," observed on December 9.

-- OP-4501was implemented by the radiation protection department, with

assistance at various points from the mechanical maintenance department. The

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inspector noted that the work order steps and the procedure steps were not well

integrated, and in some cases were duplicated. Although the individuals involved

understood their respective roles, the lack of integration of the two work

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documents could result in an inadequate charcoal sample. This issue was

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discussed with the VY maintenance superintendent. Procedure quality and

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content have been identified as areas for improvement by VY's Functional Area

Assessment and a corrective actions are planned.

increased frequency testing of the scram discharge volume vent and drain valves,

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see section E2.1.

c.

Conclusions

The surveillance activities eboarved during the period were correctly performed. However,

in one case the multiple procedures which control the standby gas treatment system

charcoal sample removal had the potential to cause errors. Activities were well controlled

and coordinated by the control room operators.

M1.3 Standbv Gas Treatment System Maintenance

a.

Insoection Scoce (62707)

The inspector observed portions of planned maintenance activities on the "B" train

standby gas treatment system (SGTS).

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b.

Observations and Findinas

On December 8, work order 98-10273 was approved for replacement of the "B" SGTS

charcoal filter test cell. In preparation for this activity, Operators tagged the "B" SGTS out

of service and placed the SGTS fan in " pull to lock." However, maintenance personnel

commenced work on the "A" SGTS train, which resulted in both trains being inoperab!e at

the sarne tirne. This problem was identified by an RP technician and was promptly

reported to the control roorn. Because the error was identified quickly, the maintenance

personnel had only begun to remove the filter train's exterior panel. Based on progression

of the work at the time of discovery, VY determined the unit would likely have performed

its safety function if required, and that there were no personnel safety issues.

VY restored the "B" SGTS to service within the TS 3.7 allowed outage time for two

inoperable SGTS subsystems (24 hrs). VY then restored frorn the inadvertent

>

maintenance on the "A" SGTS. Maintenance management conducted a work stand down

for tne department to review the incident and an Event Report was initiated.

'

The inspector determined this problem was the result of human error and that no

significant contributing causes were apparent. VY's management took immediate

corrective actions to reinforce expectations during a work stand down. A failure to follow

procedures for the conduct of maintenance is a violation of TS 6.5, Plant Operating

Procedures. This non-repetitive, licensee-identified and corrected violation is being

treated as a non-cited violation, consistent with Section Vlf.B.1 of the NRC Enforcement

Policy. (NCV 98-14-01: SGTS Maintenance Procedure implementation)

c.

Conclusions

Maintenance personnelinitiated work on the wrong standby gas treatment system filter

train and caused the entire system to be declared inoperable for a short period of time.

The error was identified by the licensee and appropriate corrective actions were initiated,

including a Maintenance department work stand down. The workers' failure to follow the

maintenance procedure is a violation of TS 6.5 and this issue was treated as a non-cited

violation.

M1.4 HPCI Low Steam Pressure Isolation Switch Not Returned To Service

a.

Inspection Scope (62707)

While per'orming a routine surveillance on the high pressure coolant injection (HPCI)

system on November 24, technicians discovered that one of four detectors for the low

steam pressure isolation function was isolated. The inspector reviewed the circumstances

surrounding this event.

b.

Observations and Findinas

While performing surveillance procedure OP-4357, HPCI Steam Line Low Pressure

Functional / Calibration, technicians identified that the instrument line valve for pressure

- .- -.

-.

-

. - - . -

_ - - .

_--. - - .-.

. ~- . - --...

-

.

.

8

switch PS-23-68D was closed and that the switch was in a tripped condition. The

pressure switch is designed to trip on low steam pressure, and inputs to a logic (one out of

two taken twice) that controls closure of the HPCI steam supply isolation valves. As such,

one half of the isolation logic was satisfied by the as-found condition. After consultation

with Operations shift supervision, the pressure detector was returned to service, the

surveillance was completed satisfactorily, and an event report was initiated.

,

The subject detector had been replaced on October 24,1998, as corrective maintenance.

W determined the detector had been inadvertently left isolated following its replacement

,

and post-maintenance testing. A root cause evaluation was in progress at the close of the

,

inspection period.

TS Table 3.2.2 requires 4 operable instrument channels for the Low HPCI Steam Supply

Pressure trip system. The TS definition of operable states, "a... component...shall be

operable or have operability when it is capable of performing its specified function (s)."

Since the pressure switch had been in a condition to initiate an isolation signal, VY

determined that it had been operable, even though its sensing line was isolated.

The inspector concluded that, although the Low HPCI Steam Supply Pressure trip system

remained capable of performing its intended safety function, the affected channel was

inoperable. The pressure switch is designed, tested, and maintained to provide an

isolation signal if steam pressure decreases to the trip setpoint. Based on the as-found

condition, the switch was not capable of its required function because the sensing line was

isolated from the HPCI steam supply. However, the inspector also concluded that, in this

case, the safety and risk significance of the isolated pressure switch was negligible.

Because the switch had been depressurized, the low steam line pressure isolation would

have functioned, if required. Also, a risk assessment by a Region I Senior Reactor

j

Analyst concluded there was an insignificant reduction in HPCI reliability as a result of the

as-found condition. Based on these findings, this failure constitutes a violation of minor

+

significance and is not subject to formal enforcement action. The inspector also noted the

W plant manager stated that a voluntary Licensee Event Report will be submitted on this

event.

l

,

in response to the technicians' finding, W initiated actions to restore the switch,

i

documented the problem in a Level 1 Event Report (98-2102), and performed an initial

review for programmatic proceduralissues that could have caused this event. Although

long term corrective actions were under evaluation by W at the close of this report period,

the inspector considered W's general response to this event reasonable.

The inspector reviewed the tagging order and procedure associated with replacement of

the pressure switch. The inspector concluded that if the administrative controls were

implemented, as written, the pressure switch would have been properly retumed to

service. The failure to follow procedures for corrective maintenance is a violation of TS , 6.5, Plant Operating Procedures. This non-repetitive, licensee-identified and corrected

violation is being treated as a Non-cited Violation, consistent with Section Vll.B.1 of the

NRC Enforcement Policy. (NCV 98-14-02: HPCI Steam Line Low Pressure isolation

Instrument isolated)

-_

-

-

,

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,-

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. _ _ _ _ _ _ . _. _ _ _ _ _ _ _ _ , _ _ . - _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . . _ . .

.

-

9

c.

Conclusions

- W identified that a pressure switch for the HPCI steam supply isolation logic had been

isolated during corrective maintenance and had not been properly returned to service.

Because the switch had been depressurized, the low steam line pressure isolation would

have functioned, if required. The failure to follow maintenance procedures was

determined to be a Non-cited Violation based on an assessment of the safety significance

of the condition and W's corrective actions.

M2

Maintenance and Material Condition of Facilities and Equipment

M2.1 Residual Heat Removal Service Water Pumo "C" Low Differential Pressure

a.

Inspection Scope (62707)

Inspection Report 50-271/98-13 discussed the November 3,1998, surveillance test failure

of the "C" RHRSW pump. After replacement of the "C" pump, W had initiated plans to

refurbish the remaining three pumps on an expedited basis. Hcwever, while performing

alternate pump testing in preparation for maintenance on the "D" RHRSW pump, the new

,

"C" RHRSW pump failed the surveillance test. The inspector reviewed W's response to

J

this problem and revised corrective action plans.

,

b.

Observations and Findinas

'

On December 3,1998, the "C" RHRSW pump failed its surveillance test acceptance

criteria for differential pressure. This pump had been replaced on November 6 and post

maintenance testing had demonstrated a 15% margin above the minimum required

differential pressure. At the time of the second failure, VY's investigation of the first failure

was not completed. On December 5, W installed a refurbished pump, completed post

maintenance testing and surveillances, and declared the "C" RHRSW pump operable. W

continued to investigate the root cause and sent the pump which failed after only a month -

in service back to the vendor for testing.

Increased frequency testing of the RHRSW pumps had been initiated after the first

surveillance test failure and the frequency was increased after the second failure. On

December 28, the "D" RHRSW pump failed the acceptance criteria for minimum

differential pressure. However, a second set of test parameters collected later that same

day indicated the pump was performing acceptably and would meet all of the inservice test

criteria. Event Report 98-2223 was initiated to capture this event in the corrective action

program. On December 31, W management approved BMO 98-44 to justify the

operability of the RHRSW pumps. The BMO addressed the limiting RHRSW pump

configuration used for the Alternate Cooling System (ACS) and concluded that with

cooling tower basin temperature of s73* F, the ACS was operable, even if the RHRSW

pump performance was degraded.

The inspector reviewed BMO 98-44 and concluded W provided an adequate basis for

i

operability of the RHRSW pumps given the heat removal capability of the system during

the winter months. The BMO evaluated both the ACS function and the post-LOCA

1

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10

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function of the RHRSW pumps. Because of the reduced deep basin temperature

assumed in the BMO (s73 F), VY performed a safety evaluation to address this change

from the system's FSAR described design condition (s105'F). The inspector verified a

sample of the procedure changes required to implement the BMO had been completed.

VY's root cause investigation was still under way at the conclusion of this inspection report

period. Several corrective actions were being pursued in parallel and the inspector

!

concluded that VY was placing priority on the long term resolution of this degraded

condition.

c.

Conclusions

The "C" residual heat removal service water pump failed inservice test acceptance criteria

I

for differential pressure on two occasions. Although immediate corrective actions restored

acceptable performance, VY developed an operability justification to address the

degradation that was observed. The operability justification was adequate and VY

management placed priority on the resolution of this degraded condition.

M2.2 Motor-Operated Valve Teraue Switch Failure

a.

Inspection Scope (62707)

On November 30,1998, the HPCI steam supply outboard isolation valve, HPCI-16, failed

to travel closed during a quarterly inservice test. The inspector observed the immediate

actions of the Operations crew and reviewed the subsequent investigation and corrective

actions by VY.

b.

Observations and Findinas

The control room operators took the actions required by TS in response to the valve's

failure by closing the redundant containment isolation valve in the penetration and

declaring the HPCI system inoperable. A conservative determination was made to report

the event as a failure of a single train safety system (reference Event Notification 35092).

This notification was later retracted after the licensee's investigation determined HPCI was

capable of performing its intended safety function, prior to being removed from service

when the penetration was isolated.

VY's investigation of the as-found condition identified that one set of contacts on the

valve's motor-operator did not have continuity. The control logic for HPCI-16 uses leaf-

style torque switch in series with the " seal-in" portion of close circuit for remote manual

operation. In (

"ast, the torque switch is bypassed in the closed direction by the circuits

for automatic isolation until the valve is 197% closed (i.e., the valve port is covered).

Based on the valve's control logic and valve being essentially open when travel stopped,

the inspector concluded the valve traveled closed while the operator held the hand switch.

After the closed limit switch provided dual control board indication, the operator released

the hand switch and the valve stopped because the seal-in circuit was not made up.

VY maintenance personnel reported that while checking continuity of the torque switch

contacts, the circuit was initially open but then made up. Close observation of the torque

.. . - -_ . . _ _ _ , _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _

__.

.

11

.

switch revealed that the second block of contacts, used for the logic of HPCI-16, was

slightly rotated in relation to the neutral position of the torque switch shaft. As a result, the

spring loaded contact fingers on one side of the contact block did not appear to have the

same contact pressure. At the end of the inspection report period, VY planned to ship the

torque switch to the vendor for evaluation.

'

The inspector reviewed VY's operability evaluation that addressed the generic implications

of the torque switch lailure. This condition was viewed as appkcable to direct current (DC)

operators since the second contact blocks were originally required on their torque

switches to minimize arcing. The inspector concluded this evaluation, coupled with an

absence of previous failures of this type, provided a good basis for VY's conclusion that no

generic operability concern exists. The inspector also noted that VY's assessment of 10

CFR 50.72 and 50.73 reporting requirements was appropriate. The isolated failure of a

containment isolation valve is not, by itself, reportable. In addition, the valve's capability to

. mitigate the consequences of a high energy line break were not impacted by the torque

j

switch failure.

'

Pending the inspectors review of the licensee's final root cause determination, evaluation

of 10 CFR 21 reportability, and maintenance rule functional failure review, this issue will

be tracked as an inspector follow-up item. (IFl 98-14-03: MOV Torque Switch Failure -

Final Resolution)

c.

Conclusions

Primary containment isolation valve HPCI-16 failed to stroke closed during an inservice

test due to the failure of the torque switch in its motor actuator. Appropriate immediate

,

actions were taken in response to the test failure, a good evaluation was made to assess

!

the generic implications of the problem and the failed torque switch was replaced.

'

Although the failure could have prevented full seating of the valve, the valve would have

closed enough to mitigate a high energy line break event. An inspector follow-up item was

- t

initiated to track NRC review of VY's final disposition of this issue.

M3

Maintenance Procedures and Documentation

M3.1 Maintenance Rule lmolementation Review

a.

Insoection Scope (62706)

10 CFR 50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear

Power Plants," (the Maintenance Rule") states in part that, in performing monitoring and

preventive maintenance activities, an assessment of the total plant equipment that is out

of service should be taken into account to determine the overall effect on performance of

safety functions. The inspector reviewed VY's process for implementing this portion of the

maintenance rule. In addition, the inspector reviewed VY's process for acquiring

performance monitoring data.

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b.

Observations and Findinas

,

l

Measures to Assess the impact of Removina Eauipment from Service

Regulatory Guide 1.160, " Monitoring the Effectiveness of Maintenance at Nuclear Power

Plants," states that NUMARC 93-01, " Industry Guideline for Monitoring the Effectiveness

of Maintenance at Nuclear Power Plants," provides methods acceptable to the NRC staff

l

for complying with the provisions of the maintenance rule. As indicated in W procedure

PP-7009, "10 CFR 50.65 Maintenance Rule Program," W utilized this guidance in

developing maintenance rule implementation plan. To assess the equipment out of

service fcr overall effect on safety functions, NUMARC 93-01 indicates that a quantitative

assessment of probabilistic risk is not required. This guidance also states that guidelines

for removing structures, systems, and components (SSCs) from service could take the

form of a matrix, a check list, a list of pre-analyzed configurations or some other utility

specific approach. It goes on to indicate that each planned maintenance activity that will

result in the removal of an SSC from service should be assessed for its impact on key

plant safety functions both during the planning and scheduling phase and prior to

authorizing removal of the SSC from service.

VY uses 1) an integrated work schedule which provides a pre-analyzed assessment of

equipment out of service and safety impact, and 2) a matrix approach for when the pre-

analyzed schedule does not address the configuration the plant would be in to support a

maintenance activity due to schedule changes and/or emergent work. The integrated

work schedule consists of a 12-week fixed schedule that was evaluated by the Safety

Assessment Group from the approach of probabilistic risk assessment (PRA).

The inspector reviewed the Safety Assessment Group's evaluation of the 12-week

schedule. Many of the systems that are managed under the 12-week schedule were not

modeled in the PRA, and therefore cannot be assessed in that context. For systems that

were modeled in the PRA, the inspector noted that the assessment of impact was based

on scheduled maintenance, rather than removal of the system from service. This was

significant in that conclusions of " insignificant impact" were often based on the fact that the

scheduled preventive maintenance activities were non-intrusive and therefore the system

was assumed to be available.

'

The inspector determined there is a potential weakness in how VY's program addresses

the inclusion of corrective maintenance during the planning phase. The implementing

procedure requires that the weekly schedule be re-assessed if the work involves a system

that was not scheduled for the particular week. However, if the corrective maintenance

involves a system scheduled for the particular week, it does not have to be reassessed,

even if the scope of work changes the equipment availability assumed in the initial PRA

based assessment. While W's overall approach is consistent with the guidance of

NUMARC 93-01, it should be recognized that inclusion of corrective maintenance in the

12-week schedule may alter the basis of the pre-analysis and may render the prior

assessment invalid. System Engineering management stated that this issue will be

reviewed for possible procedure enhancements and will be tracked under W's internal

i

'

commitment system.

_ _ .__._..._ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _

.

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13

The inspector reviewed VY's process for assessing the impact of maintenance prior to

authorizing removal of an SSC from service. VY procedure AP-0125," Plant Equipment

Control," govems this process, and reiterates the approach of using pre-analyzed

configurations and a matrix for this purpose. However, the procedural steps that discuss

use of the SSC redundancy matrix contain no binding requirements, stating only that it

provides additional guidance and identifies configurations that should be avoided.

Acauirina Performance Monitorina Data -

The two primary means of acquHng performance monitoring data are the Maintenance

Rule Out of Service Log and the Event Report process. The inspector noted that these

processes are generally effective at capturing equipment out-of-service times and

potential maintenance rule functional failures. However, an area for improvement is

capturing periods of unintended equipment unavailability. For example, a tagging error

resulted in the wrong service air compressor being removed from service; however, the

resultant unavailability was not captured in the Maintenance Rule Out of Service Log and

the Event Report was not flagged (that is, marked for routing through the Maintenance

Rule Coordinator) as a maintenance rule issue. in another example, reactor building

.

ventilation unexpectedly shut down on two occasions during this inspection period, due to

I

cold outside temperature. Event Reports were generated for both instances, but neither

was flagged as a maintenance rule issue until after the inspector discussed them with the

Maintenance Rule Coordinator.

From discussions with the work control center personnel who maintain the Maintenance

Rule Out of Service log, recording instances of unintended equipment unavailability is a

function of whether or not they are informed of it. From observations at Event Report

screening meetings, maintenance rule screening appears to concentrate on whether or

not the event constituted a maintenance rule functional failure; events relating only to

,

system reliability may not be flagged as containing maintenance rule information.

c.

Conclusions

I

'

VY's approach to the Maintenance Rule requirements for assessing the effects of out-of-

service equipment on overall safety functions is consistent with NRC-accepted guidance.

However, implementing procedures lacked positive confirmation that alternatives to the

!

pre-analyzed work had been evaluated in accordance with the program expectations.

i

VY's methods for acquiring Maintenance Rule performance monitoring data are generally

effective. However, the recording of unplanned equipment outages and the screening of

Maintenance Rule-related Event Reports are two areas where the accurate collection of

data may be challenged.

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14

,

M8

Miscellaneous Maintenance issues

M8.1 In-Office Review of LERs Related to Maintenance (90712)

An in-office review of the following LERs was performed to assess whether further NRC

actions were required. The adequacy of the overall event description, immediate actions

taken, cause determination, and corrective actions were considered during this review.

l

The following issues were closed-out based on the in-office review.

. (Closed) LER 98-017-00.01: Inadequate Design Package and Implementing Procedure

Results in Redundant Trains of the Standby Gas Treatment System with Fan Supply

Breaker Trip Setpoints Potentially Attainable with Normal Start In-Rush Current

,

!

.

l

On June 1,1998, "B" standby gas treatment (SGTS) system train failed to start on

demand. Investigation revealed that the cause was that the fan supply breaker over-

'

l

current trip setpoint was set lower than the required value. Subsequently, the "A" SGTS

l

fan supply breaker was checked and found also to have a low over-current trip setpoint.

'

Both SGTS fan supply breakers had been replaced as part of a design change that had

been installed in 1992.- VY determined that the cause of the incorrect over-current trip

setpoints was that the design change package and installation procedure had not

established the setpoint during installation. This issue was discussed in inspection report

50-271/98-08, and resulted in the issuance of two violations.

In response to this event, VY performed a safety evaluation of other safety class breakers

to ensure that over-correct setpoint control was not a generic problem. Two other

i

breakers were identified as having the wrong over-current setpoint; one of these was in a

safety class application (a standby fuel pool cooling pump) and was reported in revision 1

i

!

to the LER. In addition, procedure AP-6001, " Installation and Test and Special Test

Procedures," has been revised to strengthen the procedure pertinent to establishing

protective device settings, and procedure OP-5210, "MCC Inspection," is being revised to

reference the motor data sheet as the governing document for breaker settings. The

inspector verified that the outstanding revision is being tracked under VY's commitment

tracking system. The inspector assessed that these actions, along with the immediate

actions as discussed in inspection report 50-271/98-08, adequately addressed the

,

l

problem. Accordingly, LERs 98-017-00 and 01, and VIO 98-08-02: Design Settings Not

Translated into Installation Procedures for SGTS Breakers, are closed.

-

(Closed) LER 98-021-00: Inadequate Licensing Basis Documentation Retrievabili y

l

Results in the Failure to Meet IST Requirements for Diesel Fuel Oil Day Tank Level

Control Valves

This event occurred on August 3,1998, when it was recognized that manual ope ation of

,

l

the emergency aiesel generator (EDG) fuel oil day tank level control valves had rat been

included in the inservice testing (IST) program. Manual opeiation of these valves L,

required for continued EDG operation in the event of a loss of station / instrument air

pressure; therefore, this function is required to be verified within the IST program. In

response to this event, VY verified that the valves could be operated manually and

entered them into the IST program. This event was of minimal safety significance,

,

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because subsequent testing indicated that the valves were (and therefore, always had

been) capable of being manually operated. Failure to include manual operation of the

l

valves in the VY IST program was a violation of minor significance and is not subject to

formal enforcement action. Therefore, this LER is closed.

.

M8.2 Review of Open items (92902)

The following open item was reviewed for closure based on a review of additional

l

info!mation from VY and a sampling of the licensee's corrective actions.

(Closed) VIO 98-08-02: Design Settings Not Translated into Installation Procedures for

SGTS Breakers

The open item was reviewed and closed with the associated LER 98-017-00 discussed in

Section M8.1 of this inspection report.

'

111. Engineering

E1

Conduct of Engineering

E1.1

Imolementation of Generic Letter (GL) 96-05. " Periodic Verification of Desian-Basis

Capability of Safetv-Related Motor-Operated Valves"

a.

Insoection Scoce (Temporarv Instruction 2515/140)

Generic Letter (GL) 96-05 requested licensees to establish programs to verify

through periodic testing that safety-related motor-operated valves (MOVs) are

capable of performing their safety functions within the current licensing basis. Prior to

'

the inspection, VY responded to the recommendations of GL 96-05 in letters to the

NRC dated November 15,1996, March 13,1997, and November 3,1997.

A three-phase MOV periodic verification program developed by the Joint Owners

Group (JOG) was reviewed by the NRC staff and determined to be acceptable with

certain conditions and limitations documented in a safety evaluation report issued on

October 30,1997. In its March 13,1997 letter, VY described an alternative program

plan. This inspection evaluated VY's alternative plan to determine whether it was

l

consistent with the licensee's commitments and with the recommendations of GL 96-

l

05. The inspection was conducted through reviews of documentation and interviews

with licensee personnel. The incpectors selected a sample of MOVs considering

dynamic test availability, valve type, and risk significance to evaluate program

(

implementation. The following valves were included:

V23-15

High piessure coolant injection (HPCI) inboard containment

.

isolation (10-inch Walworth flexible wedge gate valve)

V23-16

HPCI outboard containment isolation (10-inch Walworth flexible

'

.

wedge gate valve)

1

V13-15

Reactor core isolation cooling inboard containment isolation

.

valve (3-inch Walworth solid disk gate valve)

,

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. . . - . - . - --.- - - - _ . - -

-- - .

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- V70-19A

Service water (SW) supply header cross-connect (24-inch

.

+

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Walworth solid disk gate valve)

!

V70-20

SW turbine building supply isolation (20-inch Walworth solid

-

l

disk gate valve)

b.

. Opservations and Findinas

!

Commitments to GL 96-05 (Tl 2515/140. Paraaraoh 03.01)

VY indicated that it had not committed to the JOG program because of differences

between the interim MOV static diagnostic test program being implemented at

Vermont Yankee and the interim program recommended by the JOG. The licensee

did not specify any significant objections to the other two phases of the JOG program;

i.e. the five-year dynamic test program or final periodic test program. The licensee is

suppoding the JOG program by conducting periodic dynamic tests of two service

water system valves. VY also committed to review the JOG recommendations and, if

l

necessary, the test results on which they were based, and to incorporate the results

l

of the review into its own program.

VY's attemative periodic verification plan consists of a combination of static and

dynamic diagnostic testing and periodic maintenance activities. The periodicity of

these activities is based on MOV risk significance, reliability and margin, operating

.

conditions, and the results of the performance tracking and trending program. VY

l-

intends to implement its program using the methodology described in American

Society of Mechanical Engineers (ASME) Code Case OMN-1, " Alternate Rules for

,

L

Preservice and Inservice Testing of Certain Electric Motor Operated Valve

,

Assemblies In LWR Power Plants," OM-Code-1995 Edition; Subsection ISTC.

1

GL 89-10 Lona-Term Actions (Tl 2515/140. Paraaraoh 03.02)

In Inspection Report (IR) 50-271/97-08, the NRC closed its review of the program

implemented by VY Jn response to GL 89-10, " Safety-Related Motor Operated Valve

Testing and Surveillance," based on the licensee's actions to verify the design-basis

capability of its safety-related MOVs. The IR enumerated four long-term actions in

'

support of GL 89-10 program closure, including: (1) update the current MOV program

plan to ensure that a process is in place to incorporate future test results into design

calculations; (2) reviss existing design calculations to reflect statistically derived data

from dynamically tested valves; (3) apply bounding rate of loading data from

'

previously tested globe valves to non-testable globe valves; and (4) complete the

l

Electric Power Research Institute's Performance Prediction Methodology on all

!

applicable non-tested valves and apply the resulting differential pressure thrust

values, if higher than previously calculated, in the determination of design-basis

torque switch settings.

E

in letters, dated March 2,1998 and March 30,1998, VY notified the NRC that the

'

actions had been completed. The inspectors verified through review of selected

calculations and procedures that the commitments had been met, with one exception.

Design calculation VY 98-006, " Component Level Review of Service Water (SW)

MOVs for Generic Letter 89-10," was not updated following dynamic testing of valves

..

-, . - _

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17

l

V70-19A and V70-20 in April 1998. In both cases, the new valve factors were

-

higher than those assumed in the design calculation of record. At the time of the

.

tests, the licensee informally evaluated the operabiliiy of the valves prior to returning

them to service. However, no formal operability determination, such as an

,

engineering evaluation conducted under Quality Assurance program controls, was

i

'

documented for the valves, and the need to revise the design calculations was not

'

captured in an established (e.g. engineering work or corrective action) tracking

system. Rather, the licensee informally prioritized the need to update the design

l

calculations on the basis of other work priorities and available resources.

l

The new valve factors render the current torque switch setup windows non-

'

conservative, which could result in setting the torque switches incorrectly in the

future. While W's practice of setting torque switches high in the allowable band and

l

bypassing the torque switches to the 99% closed position ameliorate the condition,

l

the current calculation of record for the valves does not reflect the actual plant

'

]

configuration. The inspectors considered the licensee's informal approach to revising

the valve design calculation to be a weakness in design control and configuration

l

management.

.

5

GL 96-05 Proaram (Tl 2515/140. Paraoraoh 03.03)

i

j.

In a November 15,1996 letter to the NRC, W stated that its GL 96-05 program,

'

. including implementing procedures and guidelines, would be established by

December 31,1997. In its November 3,1997 letter, the licensee stated the intention

to have the periodic verification program documentation completed by July 30,1998.

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While progress had been made in developing procedures and schedules, W's

program documentation was not fully developed. The licensee attributed the

condition to resource constraints and did not offer a date by which the entire program

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as described in its March 13,1997 letter to the NRC would be in place. The

inspectors' findings for specific aspects of W's GL 96-05 program were as follows:

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Scooe of MOVs included in the Prooram

The MOVs included in the periodic verification program were the same 85 valves as

those selected for the GL 89-10 program. This scope is consistent with the

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recommendations of GL 96-05.

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MOV Desian Basis

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Procedure AP 6041, " Vermont Yankee Engineering Evaluations of MOV Dynamic

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Testing and Feedback of Results into MOV Component Calculations," states that

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inputs to MOV component calculations are revised after test data has been obtained

from dynamic testing. However, there were no administrative guidelines governing

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how soon after testing the revisions are to be performed, or formal mechanisms to

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track the need to revise the calculations.

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Dearadation Rate for Potential increase in Thrust or Toraue Ooeratina Reauirements

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Dynamic test information on the potential effects of aging is needed to establish the

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rate at which the thrust required to operate gate and globe valves and the torque

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required to operate butterfly valves might increase with time. The licensee conducted

repetitive dynamic tests of service water system valves V70-19A and V70-20 in April

1998 as part of its participation in the JOG test program. However, the licensee has

not established a process to obtain information regarding degradation of the thrust

and torque operating requirements for other MOVs in its GL 96-05 program based on

appropriate periodic dynamic tests. Thus, site-specific degradation rates must be

developed and justified. The licensee indicated that its basis for establishing

degradation rates for MOV operating requirements will be evaluated and addressed

' in future program documentation.

Dearadation Rate for Potential Decrease in MOV Motor Actuator Outout

VY uses procedures OP 5219, " Diagnostic Testing of Motor Operated Valves," and

AP 6041," Vermont Yankee Engineering Evaluations of MOV Dynamic Testing and

Feedback of Results into MOV Component Calculations," to monitor potential

degradation of motor-actuator performance. Parameters affecting motor-actuator

output under static and dynamic conditions in both the opening and closing

directions, such as stem friction coefficient, motor current, load sensitive behavior,

and dynamic margin are trended. In accordance with Vermont Yankee Tracking item

. # Vendor-98013, VY is addressing new information on alternating current (AC)

powered motor actuator output provided in Limitorque Corporation Technical Update

98-01. Limitorque also noted in Supplement 1 of the technical update that new

guidance is being considered for predicting direct current (DC) powered motor

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actuator output. The licensee was proactive in incorporating industry information on

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motor actuator output by specifying the use of pullout efficiency in MOV sizing

calculations, as described in procedure AP 6038," Component Level Review of

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Vermont Yankee Motor-Operated Valves." in a sample review, the inspectors

confirmed that the licensee used pullout efficiency in its MOV calculations. VY's

review of other information contained in the Limitorque update, such as the use of

application factor and evaluation of specific MOV configurations, was ongoing, with

an assigned completion date of December 20,1998. Based on the available MOV

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capability margins, the inspectors did not identify any immediate operability concerns

resulting from the Limitorque update.

VY was developing a means to monitor motor-actuator performance degradation and

evaluating Limitorque information on motor actuator output capability. However, the

precise process for determining motor-actuator output and rates of degradation in

static and dynamic performance was not fully developed.

Periodic Test Method

VY's proposed program relies heavily on testing its MOVs under static conditions

using diagnostic equipment installed at the motor control centers (MCCs) at intervals

based primarily on risk significance. The licensee has implemented an actuator

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motor test program and undertaken a project in cooperation with CRANE MOVATS,

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incorporated to develop the equipment and software for an MCC-based diagnostic

system for DC-powered MOVs. MCC data will be verified at much longer intervals by

direct thrust / torque measurements at the valves. Current plans call for obtaining

static MCC data from high-risk MOVs every refueling outage and from medium and

low-risk MOVs every other outage. Static diagnostic testing data directly at the MOV

tentatively is scheduled for up to once every eight outages (i.e.12 years).

The inspectors compared VY's method of ranking MOVs with respect to risk

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significance with the methodology described by ti.e Boiling Water Reactor Owners

Group (BWROG) in Topical Report NEDC 32264, " Application of Probabilistic Safety

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Assessment to Generic I etter 89-10 Implementation," and the NRC safety evaluation

dated February 27,1996, accepting the BWROG methodology with certain conditions

and limitations. For example, the inspectors reviewed the licensee's use of

importance measures, consideration of common cause failure, and expert panel

oversight. The inspectors also compared the MOVs identified at Vermont Yankee as

risk significant to the composite list developed by the BWROG, and reviewed the

basis for the limited differences in those lists. The inspectors found the licensee's

risk rankings to be reasonable.

In summary, VY's methods of identifying age-related degradation affecting operating

thrust / torque requirements and motor-actuator output were still being developed at

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the time of the inspection. Development of new diagnostic equipment for remote

testing of DC-powered MOVs is a noteworthy initiative. Several aspects of the GL 96-05 program were not dully developed, such as: (1) validation of MCC-based

diagnostic equipment, particular!y for de-powered MOVs, (2) validation that the

proposed long-term test method will detect degradation under dynamic conditions, (3)

justification of test intervals that exceed ten years, and (4) finalization of test

schedules. These aspects of the licensee's periodic verification program will be

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revisited during the final review of the Vermont Yankee GL 96-05 program by the

NRC Office of Nuclear Reactor Regulation (NRR).

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MOV Performance Evaluation

As noted earlier, procedure AP 6041 provides for the evaluation of dynamic test data

and its feedback into design calculations. Procedures PP 7004, " Vermont Yankee

Nuclear Power Station Motor Operated Valve Program," and DP 0210, " Tracking and

Trending Program," provide guidance for the review and trending of MOV failures

and static and dynamic diagnostic test results every two years . Valve deficiencies

and failures also are included. The primary performance parameters that are trended

are unseating thrust, thrust of close control switch trip (CST), total close thrust,

average running thrust and motor current, stem friction at close CST and backseating

thrust (if any). Significant differences from previous tests are noted and additional

parameters may be chosen for review as necessary. For the selected valves, the

inspectors reviewed the most current (1996) diagnostic testing trend report. The

report was comprehensive and contained a considerable amount of test data for each

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valve. No performance anomalies or adverse trends over the previous two years

were noted, and causes of significant performance differences were evaluated and

discussed.

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MOV Test interval

VY is considering a static diagnostic test program that combines frequent data

acquisition from the MCCs supplemented periodically by comparison with data taken

directly at the MOVs to assure design-basis capability between scheduled tests. As

an alternative to committing to the JOG program, the licensee must establish its own

long-term periodic test methods and intervals for establishing thrust / torque

degradation rates for each MOV in order to satisfy GL 96-05. The licensee plans to

monitor potential degradation of motor-actuator output through static and dynamic

testing. VY's technical justifications for its long-term MOV periodic test methods and

intervals were still under development. When the GL 96-05 program documentation

is prepared, NRR will review the licensee's methods for monitoring parameters to

ensure adequate capability between normally scheduled tests.

c.

Conclusions

Within the scope of the review performed on site, the inspectors concluded that the

actions to date by VY concerning periodic verification of long term MOV capability

were acceptable. The NRC Office of Nuclear Reactor Regulation will use this

information in preparing a safety evaluation on VY's response to GL 96-05. Positive

aspects of the periodic verification program were observed, including: (1)

development of more efficient test techniques,(2) implementation of a motor test

program, and (3) an aggressive motor-actuator lubrication and refurbishment

schedule. However, several aspects of the periodic verification program, such as

program documentation and MOV performance degradation rates, as yet were

incomplete. Design-basis thrust calculations for two MOVs were not revised to reflect

dynamic test information, resulting in the calculations not reflecting the actual plant

configuration. This condition was an example of poor configuration management.

E2

Engineering Support of Facilities and Equipment

E2.1

Scram Discharae Volume Drain Valve Failures

a.

Insoection Scope (62707,37551)

On December 6, containment isolatici vdve CRD-33D, the outboard drain isolation valve

for the south scram discharge volume 60V), failed to shut during a quarterly inservice

test. Following the repair of CRD-33D, VY initiated increased frequency testing for all four

SDV drain valves, pending a final root cause determination. On December 11, a second

SDV drain valve, CRD-33B, failed to shut within the required time and was declared

inoperable. The inspector examined VY's corrective actions and cause determination for

these failures.

b.

Observations and Findinas

The SDV is considered an extension of the primary containment and TS 3.7.2 requires the

SDV drain line to bc isolated if one of its automatic isolation valves is inoperable.

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However, because of minor HCU valve leakage, water accumulated in the south SDV and

required compensatory operator actions to keep the SDV drained.

The inspector reviewed VY's immediate actions to address the individual valve failures.

Compensatory actions were initially adequate and later were enhanced after W was self-

critical of the process used to implement the compensatory actions. There are three main

control panel indications that allow operators to monitor the SDV level when the drains are

isolated. This was the situation while the corrective actions were being planned and

implemented. The inspector concluded that the operators were provided with adequate

guidance and information to implement the compensatory measures.

The vent and drain valves for the north and south SDVs were modified during the 1998

refueling outage to resolve a long standing problem of corrosion products impacting the

valves' ability to meet pressure test acceptance criteria. VY selected new ball valve and

air operator pairs from BW/IP International based on the performance of similar ball valve

configurations in other plant applications. The new valves and actuators were installed

under an engineering design change, EDCR 97-410 with a 50.59 safety evaluation.

The inspector reviewed EDCR 97-410 and made the following observations:

The procurement specification provides no details regarding the cundition of the

process flow (i.e., intermittent water and air with corrosion products).

The safety evaluation states the new drain valves require a maximum torque of s

360 in-lbs (30 ft-lbs) to actuate and the vendor supplied valve data sheet shows a

required stem torque of 110 ft-Ibs.

W did not independently evaluate design aspects of matching the valve torque

requirements to the actuator capability and did not require documentation of this

activity in the purchase specification. Calculations and certification for other critical

design information such as closure time, seismic capability and material

traceability were required from the vendor.

The inspector concluded that the design change process failed to identify the

inadequate valve to actuator sizing for this application, pricr to installation of the

modification and subsequent failures while in service. Additionalinformation from

W is necessary to assess whether the process or its implementation was the

cause of this failure. Also, additional information is necessary for the inspector to

assess W's review of 10 CFR 21, Reporting of Defects and Noncompliance,

applicability.

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On December 14, W reported the problems observed with the SDV drain line

isolation valves based on the common mode failure mechanism and the reporting

requirements of 10 CFR 50.72. As such, a Licensee Event Report is expected

based on the parallel reporting requirements of 10 CFR 50.73.

10 CFR 50 Appendix B, Criterion Ill, " Design Control," requires, in part that

measures be established for the selection and review for suitability of application of

equipment that are essential to the safety related functions of systems and

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components. In accordance with the guidancc provided in NRC Enforcement

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Guidance Memorandum 98-006, this issue, which may represent a violation of

NRC requirements will remain open for a reasonable time to all the licensee to

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develop its corrective actions. (eel 98-14-04: Design Control For SDV Valve

Modification)

c.

Conclusions

The inservice test failures of two scram discharge volume drain valves led to the

identification of problems with the new valve / actuator design installed during the 1998

refueling outage. Errors were identified in the safety evaluation and there is a lack of

design information from the vendor. in accordance with NRC guidance, this issue, which

may represent a violation of NRC requirements will remain open for a reasonable time to

allow the licensee to develop its corrective actions.

E8

Miscellaneous Engineering issues

E8.1

Review of Open items (92903)

The following open items were reviewed for closure based on a review of additional

information from VY and a sampling of the licensee's corrective actions.

iClosed) URI 98-80-08: Complete Review of February 25,1998, Submittal

This item was opened because a VY submittal dated February 25,1998, (BVY 98-22)

incorrectly identified a dose criteria associated with the designation of structures, systems,

or components as Safety Class 3. A VY contractor study, dated February 6, identified the

dose criteria in the VY Safety Class Manual (reflected in the February 25 VY letter)

conflicted with the ANS 22 dose criteria required by the NRC-approved Operational

Quality Assurance Program.

Based on a review of VY's February 25,1998, letter, the inspector determined that the

ANS 22 dose criteria for classification of equipment was confused with the licensing basis

dose limit for abnormal operational occurrences listed in FSAR Chapter 14.2. In a letter

dated April 10,1998, VY retracted its February 25,1998, submittal to the NRC regarding

Revision 28 of the Operational Quality Assurance Program. VY initiated an Event Report

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to evaluate the impact of the dose criteria misunderstanding and appropriate initial actions

were taken. Violation 50-271/98-80-07 was previously issued for the licensee's reduction

in Quality Assurance commitments without prior NRC approval. The incorrect information

was retracted by VY and therefore, the inspector concluded that no violation of NRC

requirements existed. This item is closed.

E8.2

Review of VY Cycle 19 Ooere*jna Report

By letter dated December 1,1998, the licensee submitted the Vermont Yankee Cycle 19

Operating Report in accordance with 10CFR50.59(b)(2) and 10 CFR50.4. This report

contains a brief description of the changes, tests, and experiments, including a summary

of the safety evaluation of each, which were conducted between November 2,1996 and

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June 1,1998. An in-house review of this report was conducted by the NRR Project

Manager. The report described the changes, tests, and experiments in sufficient detail to

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support the licensee's conclusion that the changes did not involve unreviewed safety

questions. No concerns were identified during the review.

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IV. Plant Support

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EP Procedures and Documentation

P3.1

in-Office Review of Emeraency Plan Imolementina Procedures (82701)

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The inspector reviewed recent changes the licensee made to its Emergency Plan

Implementing Procedures in the NRC Region I office.

Based on the licensee's determination that the changes do not decrease the overall

effectiveness of its emergency plan, no prior NRC approval is required in accordance with

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10 CFR 50.54(q). After a limited, in-office review of the changes, the inspector concluded

that these changes were made in accordance with the provisions of 50.54(q).

Implementation of these changes will be subject to future on-site inspection effort to

confirm that the changes have not decreased the effectiveness of the licensee's

emergency plan. A list of the changes reviewed are included as an attachment to this

report.

V. Management Meetings

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Exit Meeting Summary

The resident inspectors met with licensee representatives periodically throughout the

inspection and following the conclusion of the inspection on January 26,1998. At that

time, the purpose and scope of the inspection were reviewed, and the preliminary findings

were presented, in addition, visiting inspector debriefed with the licensee prior to leaving

the cite. The licensee acknowledged the preliminary inspection findings.

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The Inspector asked the licensee whether any material examined during the inspection

should be considered proprietary. No proprietary information was identified.

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ATTACHMENT 1

List of Acronyms Used

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ACS

Alternate Cooling System

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ANS

American National Standard

ARM

Area Radiation Monitor -

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.BMO

Basis for Maintaining Operation

BWROG

Boiling Water Reactor Owners Group

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CFR

Code of Federal Regulations

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. CIV

Containment isolation Valve'

CRD ~

, Control Rod Drive

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CS

Core Spray

EDCR

Engineering Design Change Request

EDG

Emergency Diesel Generator..

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ER

' Event Report

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FME-

Foreign Material Exclusion

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FSAR

- Final Safety Analysis Report

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GL~

Generic Letter

.HCU

Hydraulic Control Unit

HPCI

High pressure coci

.Mjection

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inspector Follow-up R,m .

IR(s)

Inspection reports (s)

IST -

Inservice Test

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LCO

Limiting Condition for Operation

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LER

Licensee Event Report

LOCA

Loss of Coolant Accident

MCC

Motor Control Center

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MOV-

Motor Operated Valve

NRC

Nuclear Regulatory Commission

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NRR

NRC Office of Nuclear Reactor Regulation

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NUMARC Nuclear Management and Resources Council

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PRA

Probabilistic Risk Assessment

QA .

Quality Assurance

QC -

Quality Control

RCIC

Reactor core isolation cooling

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RHR

Residual Heat Removal

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RHRSW

Residual Heat Removal Service Water

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.RP

Radiation Protection

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SDC'

Shutdown Cooling

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lSDV

Scram Discharge Volume

SGTS

Standby Gas Treatment System

SOV(s) -

Solenoid-operated valve (s)

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List of Acronyms Used

!SSC

Structure, System, or Component

SW

Service water

TS

Technical Specifications

URI

Unresolved item

VDC'

Volts Direct Current

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VIO

Violation

VY

Vermont Yankee

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ATTACHMENT 2

Items Opened, Closed, or Discussed

Opened

IFl 98-14-03: MOV Torque Switch Failure - Final Resolution (page 11)

eel 98-14-04: Design Control For SDV Valve Modification (page 22)

Closed

LER 98-020-01: Inadequate Equipment Control Practices Result in Two Mispositioned Isolation

Valves Allowing Degradation of Primary Containment integrity (page 3)

LER 98-019-00: Off-Normal System Alignment Following a Plant Trip Which involved the Loss of

a Reactor Water Recirculation System Pump and Reactor Water Thermal Stratification

Results in a Spurious Shutdown Cooling isolation (page 4)

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LER 98-017-00,01: Inadequate Design Package and implementing Procedure Results in

Redundant Trains of the Standby Gas Treatment System with Fan Supply Breaker Trip

Setpoints Potentially Attainable with Normal Start in-Rush Current (page 14)

LER 98-021-00: Inadequate Licensing Basis Documentation Retrievability Results in the Failure

to Meet IST Requirements for Diesel Fuel Oil Day Tank Level Control Valves (page 14)

VIO 98-08-02: Design Settings Not Translated into Installation Procedures for SGTS Breakers

(page 15)

URI 98-80-08: Complete Review of February 25,1998, Submittal (page 22)

Non-cited Violations

NCV 98-14-01: SGTS Maintenance Procedure Implementation (page 7)

NCV 98-14-02: HPCI Steam Line Low Pressure Isolation Instrument Isolated (page 8)

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ATTACHMENT 3

Emergency Response Plan and implementing Procedures Reviewed

Procedure

Title

Rev.No.

OP-3504

Emergency Communications

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98-413

OP-3505

Emergency Preparedness Exercises and Drills

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OP-3506

Emergency Equipment Readiness Check

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OP-3508

On-Site Medical Emergency Procedure

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OP-3513

Evaluation of Off-Site Radiological Conditions

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98-372

OP-3524

Emergency Actions to Ensure Initial Accountability and Security

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Response

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OP-3531

Emergency Call-in Method

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98-414

OP-3532

Emergency Preparedness Organization

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OP-3534

Post Accident Sampling of Plant Stack Gaseous Releases

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