IR 05000271/1998014

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

ML20202G278
Person / Time
Site:	Vermont Yankee
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)

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i U.S. NUCLEAR REGULATORY COMMISSION -I REGION I j 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

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Dates: November 22,1998 - January 4,1999 Inspectors: Brian J. McDermott, Senior Resident inspector Edward C. Knutson, Resident inspector William A. Maier, Emergency Preparedness Specialist

. Richard P. Croteau, Project Manager, NR3 Douglas A. Dempsey, Reactor Engineer Thomas G. Scarbrough, Senior Mechanical Engineer, NRR l

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Approved by: Clifford J. Anderson, Chief, Projects Branch 5 Division of Reactor Projects l

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. 1 EXECUTIVE SUMMARY l

Vermont Yankee Nuclear Power Station 1 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 j 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)

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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 l 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) l l

Maintenance i l

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

charcoal sample removal had the potential to cause errors. Activities were well controlled l 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. l Because the switch had been depressurized, the low steam line pressure isolation would i 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 ]

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 i 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)

t * W'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. (Section M3.1)  !

Enaineerina

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

operated valves (MOVs) 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 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 iii

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- Executive Summary (cont'd)

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

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TABLE OF CONTENTS

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l EXECUTIVE SU MMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 -

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

- Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1. Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 Conduct of 0pe rations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 O1.1 Area Radiation Monitor Power Supply Failure . . . . . . . . . . . . . . . . . . . 1

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j 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . , . 2

03.1 Administrative Control of Manual Containment Isolation Valves . . . . . . 2 08= Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 i 08.1 In-Office Review of LERs Related to Operations . . . . . . . . . . . . . . . . . 3 i

l l . Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M1 ' Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 L

M1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l 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 l M2.1 ' Residual Heat Removal Service Water Pump "C" Low Differential

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Pressu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M2.2 Motor-Operated Valve Torque Switch Failure . . . . . . . . . . . . . . . . . . . 10 I M3 Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 11 '

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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! E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 j E1.1 Implementation of Generic Letter (GL) 96-05, " Periodic Verification of i Design-Basis Capability of Safety-Related Motor-Operated Valves" . . 15 i E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 20 E2.1 Scram Discharge Volume Drain Valve Failures . . . . . . . . . . . . . . . . . 20 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 ,

E8.1 Review of Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 J E8.2 Review of VY Cycle 19 Operating Report . . . . . . . . . . . . . . . . . . . . . . 22 IV. Plant Support . . . . . . . . . . . . . . . . . . . . .................................... 23 i 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 X1 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 vi

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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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01 . 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. .

The two ARMS were declared inoperable at 7:30 a.m. on November 24. Technical

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Specification 3.2, " Protective Instrument Systems," Table 3.2.3, requires that the reactor 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 l 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 -

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 /> !

after the problem was identified. >

c. Conclusions '

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

l prompt and adequately resolved the degraded condition.

03 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

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

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

, 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 4 integrity.

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 q approval.

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

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

a 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.

i 08 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

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

(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 i 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 97-11585, observed on December 8 and 12.

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l -- Observed preliminary disassembly to support the pump casing lift the following day, i- Work was being pc rformed in a contaminated area with dedicated RP support and full i time foreman supavision. No deficiencies were noted. l

-- On December 12, observed preparations to land the pump casing. Level 3 FME l controls were in effect and the material accountability log was being properly maintained.

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

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

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 j activities.

c. Conclusions The maintenance activities observed during this period were performed well. Workers l demonstrated appropriate radiological control and foreign material exclusion control L 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. i

-- 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, i

" Filter Tosting," observed on December 9.

-- OP-4501was implemented by the radiation protection department, with i assistance at various points from the mechanical maintenance department. The j inspector noted that the work order steps and the procedure steps were not well i integrated, and in some cases were duplicated. Although the individuals involved i understood their respective roles, the lack of integration of the two work i documents could result in an inadequate charcoal sample. This issue was )

discussed with the VY maintenance superintendent. Procedure quality and i content have been identified as areas for improvement by VY's Functional Area Assessment and a corrective actions are planned.

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increased frequency testing of the scram discharge volume vent and drain valves, see section E2.1.

c. Conclusions l

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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

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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.

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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

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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

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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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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 J

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

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this problem and revised corrective action plans.

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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 - 1 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

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l 10 l 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

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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.

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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

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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 !

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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 -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 l

Measures to Assess the impact of Removina Eauipment from Service l

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.

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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 i

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

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commitment system.

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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

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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. l c. Conclusions

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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. l 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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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.

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

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. (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

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

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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.

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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

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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.

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(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

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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.

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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

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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

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. 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)

. V13-15 Reactor core isolation cooling inboard containment isolation valve (3-inch Walworth solid disk gate valve)

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+ - V70-19A Service water (SW) supply header cross-connect (24-inch .

l Walworth solid disk gate valve)

! - V70-20 SW turbine building supply isolation (20-inch Walworth solid l

disk gate valve)

b. . Opservations and Findinas

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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

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conditions, and the results of the performance tracking and trending program. VY l- intends to implement its program using the methodology described in American

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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. l

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

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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

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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

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

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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 ,

i engineering evaluation conducted under Quality Assurance program controls, was '

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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.

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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.

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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,

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. 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

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

f 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.

j: MOV Desian Basis I

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Procedure AP 6041, " Vermont Yankee Engineering Evaluations of MOV Dynamic Testing and Feedback of Results into MOV Component Calculations," states that j inputs to MOV component calculations are revised after test data has been obtained

from dynamic testing. However, there were no administrative guidelines governing l 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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r i 18 i Dearadation Rate for Potential increase in Thrust or Toraue Ooeratina Reauirements l

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

. rate at which the thrust required to operate gate and globe valves and the torque 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) i 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 j

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actuator output. The licensee was proactive in incorporating industry information on motor actuator output by specifying the use of pullout efficiency in MOV sizing calculations, as described in procedure AP 6038," Component Level Review of '

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

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).

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The inspectors compared VY's method of ranking MOVs with respect to risk significance with the methodology described by ti.e Boiling Water Reactor Owners

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

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

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thrust / torque requirements and motor-actuator output were still being developed at 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 i revisited during the final review of the Vermont Yankee GL 96-05 program by the l NRC Office of Nuclear Reactor Regulation (NRR).  !

i MOV Performance Evaluation l 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

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report was comprehensive and contained a considerable amount of test data for each 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, i applicability.

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 j Guidance Memorandum 98-006, this issue, which may represent a violation of l NRC requirements will remain open for a reasonable time to all the licensee to 1 develop its corrective actions. (eel 98-14-04: Design Control For SDV Valve i 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 I 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 '

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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23 I 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 l 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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P3 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 l 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 i 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 l report.

l V. Management Meetings-X1 Exit Meeting Summary l

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. j 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 i ACS Alternate Cooling System i ANS American National Standard ARM Area Radiation Monitor - J

.BMO Basis for Maintaining Operation !

BWROG Boiling Water Reactor Owners Group ]

CFR Code of Federal Regulations j

. CIV Containment isolation Valve'  :

CRD ~ , Control Rod Drive i CS Core Spray  !

EDCR Engineering Design Change Request EDG Emergency Diesel Generator.. j ER ' Event Report i FME- Foreign Material Exclusion  !

FSAR - Final Safety Analysis Report l' GL~ Generic Letter

.HCU Hydraulic Control Unit

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HPCI High pressure coci .Mjection

'IFl inspector Follow-up R,m .

IR(s) Inspection reports (s)

IST - Inservice Test <

LCO Limiting Condition for Operation I i LER Licensee Event Report  !

LOCA Loss of Coolant Accident MCC Motor Control Center ,

j MOV- Motor Operated Valve  ;

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NRC Nuclear Regulatory Commission NRR NRC Office of Nuclear Reactor Regulation l l NUMARC Nuclear Management and Resources Council l l' PRA Probabilistic Risk Assessment QA . Quality Assurance  !

QC - Quality Control RCIC Reactor core isolation cooling i I RHR Residual Heat Removal l l- RHRSW Residual Heat Removal Service Water l' .RP Radiation Protection l: SDC' Shutdown Cooling ,

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I lSDV Scram Discharge Volume SGTS Standby Gas Treatment System SOV(s) - Solenoid-operated valve (s)

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- Attachment 1 (continued) '

List of Acronyms Used

!SSC Structure, System, or Component SW Service water TS Technical Specifications URI Unresolved item

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VDC' Volts Direct Current 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) -

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 30,D1 -

98-413 OP-3505 Emergency Preparedness Exercises and Drills 21 OP-3506 Emergency Equipment Readiness Check 36 OP-3508 On-Site Medical Emergency Procedure 21 ,

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OP-3513 Evaluation of Off-Site Radiological Conditions 19,DI l 98-372 i OP-3524 Emergency Actions to Ensure Initial Accountability and Security 15 Response

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OP-3531 Emergency Call-in Method 11,DI 98-414 OP-3532 Emergency Preparedness Organization 7 OP-3534 Post Accident Sampling of Plant Stack Gaseous Releases 2 l

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