IR 05000271/1998012

Insp Rept 50-271/98-12 on 980830-1010.No Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML20195D580
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/10/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20195D577	List: IR 05000271/1998012 ML20195D571
References
50-271-98-12, NUDOCS 9811180100
Download: ML20195D580 (35)

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i U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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i-Docket No.

50-271

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Licensee No.

DPR-28 Report No.

98-12 r

Licensee:

Vermont Yankee Nuclear Power Corporation l

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> Facility:

Vermont Yankee Nuclear Power. Station L'

Location:

Vernon, Vermont Dates:

August 30 - October 10,1998

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Inspectors:

Brian J.' McDermott, Senior Resident inspector Edward C. Knutson, Resident inspector Todd H. Fish, Operations Engineer Paul R. Frechette, Physical Security inspector George W. Morris, Senior Reactor Engineer David G. Cullison, Reactor Engineer Robert J. Summers, Project Engineer

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Approved by:

Curtis J. Cowgill, Ill, Chief,' Projects Branch 5 Division of Reactor Projects

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EXECUTIVE SUMMARY l

Vermont Yankee Nuclear Power Station NRC Inspection Report 50-271/98-12 This integrated inspection report includes aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six week period of routine Resident i

inspector activities and announced inspections by a Physical Security inspector and an Operations Engineer.

Ooerations A licensed operator requalification program inspection found that VY developed

acceptable operating exams which included an appropriate sampling of knowledge and performance areas. The training staff objectively assessed operator performance for one crew and the inspectors agreed with the VY evaluators'

conclusion regarding the results of the observed examination. (Section 05.1)

i Maintenance Maintenance activities were adequately controlled by the generalinstructions

provided in work orders and procedures. Based on the activities observed, VY relies heavily on the knowledge and training of its workforce since the written instructions contain little detail. Post maintenance testing and surveillance activities demonstrated that the equipment was properly restored and operable. (Section M 1.1 )

VY responded promptly to eliminate an NRC-identified mechanical interference

between the solenoid for a directional control valve and the in!et scram valve on one hydraulic control unit (HCU). VY determined that the solenoid could have been damaged by operation of the scram valve but, the operability of the HCU was not affected. VY developed appropriate long term corrective action. (Section M2.1)

Enaineerina VY engineers identified a scenario which could result in post-LOCA pressurization of e

the secondary containment and this apparently was not evaluated in the original licensing basis. Preliminary VY evaluations indicate that pressurization would occur, i

but the analyses were not finalized because the VY licensing department concluded that secondary containment pressurization was not part of VY's original licensing basis. This issue requires further NRC evaluation and will be tracked as an i

inspector follow-up item. (Section E1.1)

Plant Support

Performance testing of the perimeter intrusion detection system (PIDS), by the

licensee and NRC program office personnel, resulted in the appropriate intrusion alarms being generated in all zones tested. In addition, the licensee demonstrated

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- proper searches of packages entering the protected area through the access control l

point. (Section S8)

VY initiatives identified several fire barrier penetrations in the plant that were not

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configured or constructed as designed. These findings were reported in Licensee l

Event Reports (LERs) and appropriate programmatic corrective actions are being taken. Although a fire barrier existed in each instance, their 3-hour rating was

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degraded. These deficiencies, individually or taken together, did not cause a significant degradation of the overall fire protection capabilities. (Section F2.1)

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TABLE OF CONTENTS EX E C UTIV E S U M M A RY.............................................. ii

l TA B L E O F CO N T E N TS.............................................. iv

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l Summary of Plant Status

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l. Operations

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Conduct of Operations................................... 1 01.1 Observation of Routine Shift Operations................... 1

Operator Training and Qualification........................... 1 05.1 Licensed Operator Requalification Training (LORT) Program..... 1

Miscellaneous Operations issues............................. 3 08.1 Review of Open items

...............................3 08.2 in-Office Review of LERs Related to Operations.............. 3 II. M a int e n a n c e................................................... 4 M1 Conduct of Maintenance................................... 4

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M 1.1 Maintenance Observations............................ 4 M 1.2 Surveillance Observations............................. 5 M2 Maintenance and Material Condition of Facilities and Equipment....... 5 M 2.1 Directional Control Valve / Scram Valve Mechanical Interference.. 5 M8 Miscellaneous Maintenance issues............................ 6 M8.1 Review of Open Items............................... 6 M8.2 In-office Review of LERs Related to Maintenance and Surveillance

.7 111. E n g in e e ri n g................................................... 9 E1 Conduct of Engineering.................................... 9 E1.1 (Opened) IFl 98-12-02: Potential for Post-LOCA Reactor Building Pr e s s u riz at io n..................................... 9 E8 Miscellaneous Engineering issues............................

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E8.1 Review of Open Items

..............................10 E8.2 In-office Review of LERs Related to Engineering............

E8.3 Review of Selected Event Reports......................

IV. Plant Support

................................................21 S8 Miscellaneous Security and Safeguards issues

..................21 S8.1 (Closed) VIO 9 8 -0 5 -01.....................

........21 S8.2 (Closed) VIO 9 8 -0 5 -0 2.............................. 2 2 S8.3 (Closed) VIO 9 8 -0 4-0 6............................. 23 F2 Status of Fire Protection Facilities and Equipment................23 F2.1 Fire Barrier Penetration Seals.......................... 23 F8 Miscellaneous Fire Protections issues.................

.......25 F8.1 in-office Review of LERs Related to Fire Protection.......... 25 F8.2 Review of Open items

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a V. Management Meetings

..........................................26 X1 Exit Meeting Summ ary................................... 2 6

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- NRC/ Vermont Yankee Public Meeting............,............. 27 ITEMS OPENED, CLOSED, AND DISCUSSED

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LIST O F ACRO NYM S USED.......................................... 30

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Report Details Summarv of Plant Status Throughout this report period, Vermont Yankee (VY) was investigating problems with the main turbine control system that had resulted in control valve oscillations near 100 percent power. As a compensatory measure, power was reduced to below the level at which the oscillations wem observed. Although linkage problems associated with the electronic pressure regulator (EPR) were identified and corrected, the control valve oscillations recurred on September 9,1998 during main turbine stop valve testing. VY management instituted an administrative power limit of 95 percent in response to this occurrence, pending further investigation. VY identified several potential adjustments and one diode failure that may have contributed to the control system problem. At the close of this report period, a plan for post maintenance testing had been developed and was awaiting final approval and implementation.

l. Operations

Conduct of Operations'

01.1 Observation of Routine Shift Ooerations (71707)

The inspectors toured the control room to assess the conduct of activities, verify safety system alignments, and verify compliance with Technical Specification (TS)

requirements. Event Reports used to document plant deficiencies were reviewed, and discussed with shift supervision, to evaluate both the equipment condition l

discussed and the licensee's initial response to the issue. No significant deficiencies

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were noted.

Operator Training and Qualification 05.1 Licensed Ooerator Reaualification Trainina (LORT) Proaram a.

Insoection Scope (71001)

Examiners assessed the licensed operator requalification training (LORT) program using selected portions of NRC Inspection Procedure 71001," Licensed Operator Requalification Program Evaluation." The examiners observed performance and evaluations of a shift crew during the annual operating examination. During the scenario evaluations, the resident inspctor assisted the examiners.

' Topical headings such as O1, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic...

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Observations and Findinas Event Review The examiners reviewed Licensee Event Reports (LERs), resident inspector reports, and discussed recent operator performance with the resident staff. The inspectors concluded that the identified operator performance issues were not indicative of

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problems with the LORT program. The examiners noted the program contained on-going training topics to reinforce management's expectations regarding performance standards.

Exam Materials The examiner reviewed the annual operating test. The materialincluded three dynamic simulator scenarios and six job performance measures (JPMs). The test items required the operators to demonstrate an understanding of, and the ability to perform, a representative sample of the items described for operating tests in 10

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CFR 55.45. Also, the materials met guidelines for quality established in NUREG

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1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision b. The examiner specifically noted that senior reactor operators were tested on their knowledge of the emergency plan, including whether the plan should be executed and the reactor operators' duties under the plan. Based on a sampling of the scenario bank, the examiner noted that test items were in place to evaluate operators on their ability to operate the facility between shutdown and designated power levels.

Exam Administration The inspectors observed the training staff and Operations Manager administer two dynamic simulator examinations to one operating crew. The training staff thoroughly evaluated the operators' performance against clear and objective standards. The resident inspector noted that crew performance essentially mirrored that of the operators in the plant's control room. The examiners agreed with the training staff's conclusion that all observed individuals passed the scenario exams.

Security measures were appropriate during the administration of the scenarios and JPMs. Examiners also reviewed the exams administered to all operators (six operating crews; two staff crews) and noted that the overlap for scenario and JPM test items between the crews was very low, c.

Conclusions The test materials contained a sample of the items required by 10 CFR 55.45 and met the guidelines for quality described in NUREG 1021. The training staff adequately administered the operating exam and thoroughly evaluated operator performance. The examiners agreed with the VY training staff's conclusion that all observed individuals passed the examinations.

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08 Miscellaneous Operations issues 08.1 Review of Ooen Items (92901)

The following open item was reviewed for closure based on a sampling of the licensee's corrective actions.

(Closed) VIO 98-09-01: Plant Operating Procedures This violation addressed examples of inadequate procedures identified during a Special Team inspection chartered to review the June 9,1998 reactor scram.

Procedures governing operator response to reactor high water level and for restoring electrical systems did not adequately direct or alert operators to certain system responses while executing steps of those procedures. In response to the violation, the facility staff determined that the procedures were too general and did not contain appropriate precautions about the automatic restart potential for equipment.

Subsequently, the staff revised the respective procedures to address these deficiencies. The inspector reviewed the procedure revisions, determined they were, appropriate, and confirmed the revisions had been implemented. This violation is

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closed.

08.2 in-Office Review of LERs Related to Operations (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 i

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

(Closed) LER 97-023-00: A Componert Failure in the Main Generator Protection Circuitry Results in a Reactor Scram This event was documented in inspection report 50-271/97-11,and resulted in a violation being issued for f ailure to review and approve switching orders. An electrical transient that occurred while conducting a switching operation in the main switchyard was the initiating event for this scram. VY's corrective actions in response to this violation were reviewed in inspection report 50-271/98-11,and the violation was closed. VY determined that the root cause of the event was failure of a portion of the generator runback circuit which caused the runback to go to completion (rather than stopping when the condition requiring the runback cleared),

thereby producing the transient (due to reduced feedwater heating) which caused the scram.

The inspector determined that the LER adequately described the event and that the root and contributing cause determinations were reasonable. Immediate corrective action included removal of the current-to-flow comparitor from the main generator protective circuitry, thus eliminating the root cause of the event. Long term corrective action was to evaluate alternatives to the current-to-flow comparitor

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circuitry, as well as to evaluate the possibility of removal of the turbine runback circuitry. Based on discussions with engineering personnel, a modification to remove the runback circuitry is now tentatively planned for the 1999 refueling outage. The automatic generator protective feature will be replaced by operator action, initiated on the basis of stator cooling water temperature, pressure, and flow. These indications will be installed in the control room as part of an engineering design change. The inspector concluded that these actions, along with those taken in response to the violation, adequately addressed the causes of the event.

II. Maintenance M1 Conduct of Maintenance i

M 1.1 Maintenance Observations

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Inspection Scoce (62707)

j The inspector observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and TS

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requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to

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ensure that equipment operability was verified upon completion of post maintenance testing.

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Observations and Findinas Field observations were made during portions of the following activities:

WO 98-05644-00, Emergency Diesel Generator Preventive Maintenance,

performed September 21 - 24. Maintenance personnel experienced difficulty in disassembly of the vertical drive coupling and later determined that the disassembly sequence was initially incorrect. No equipment damage was observed as a result of the disassembly problems.

WO 98-05644-00, Emergency Diesel Generator Air Start Check Valve

Disassembly and Inspection, performed September 22. Maintenance l

personnel were knowledgeable regarding the task and vendor manual

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guidance.

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WO 98-05648-01, Safety Related Electrical Breaker inspection and Replacement performed on September 22. Maintenance personnel initially overlooked an insulator during reassembly and installation of the replacement molded case circuit breaker but the oversight was corrected prior to the post maintenance testing.

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Conclusions Maintenance activities were adequately controlled by the general instructions provided in work orders and procedures. Based on the activities observed, the inspector concluded VY relies heavily on the knowledge and training of its workforce since the written instructions are not detailed. Post maintenance testing and surveillance activities demonstrated that the equipment was properly restored and operable.

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M1.2 Surveillance Observations a.

Inspection Scope (61726)'

The inspector observed portions of surveillance tests 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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_ Observations and Findinus Portions of the following surveillance activities were observed by the inspectors:

Inservice Testing of the Reactor Building Closed Cooling Water system

l performed in accordance with OP-4182 was observed on September 24. No deficiencies were noted.

Core Spray pump surveillance, performed in accordance with OP-4123 was

observed on October 6. The inspector observed testing of the "A" pump from the control room, and of the "B" pump locally. No deficiencies were noted, i

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Surveillance tests observed during this inspection were performed in accordance l

with plant procedures. Technical Specification requirements for equipment allowed l

outage time were met and auxiliary operators were effective in coordinating and i

communicating the test activities with the control room.

M2 Meintenance and Material Condition of Facilities and Equipment M 2.1 pirectional Control Valve / Scram Valve Mechanical Interference a.

Inspection Scope (62707)

On September 9, the inspector observed a mechanical interference problem on one j

of the hydraulic control units (HCUs). Specifically, the electrical connector for a i

directional control valve was rotated and positioned directly over the position i

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indicator for a scram valve. Because the scram valve stem moves upward when the i

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valve opens, it would have hit the solenoid's electrical connection, were a scram to occur. The inspector was concerned that this mechanicalinterference might slow the opening time of the scram valve and adversely affect the scram time of the associated control rod, b.

Observations and Findinas The inspector informed VY of the condition, and VY responded by rotating the solenoid to clear the interference. The remaining 88 HCUs were inspected and no other examples were identified. Although this potential interference issue was identified in a 1973 General Electric Service Information Letter (SIL-3), the inspector determined this information pre-dated VY commitments in response to NUREG 0737 Item I.C.5, Feedback of Operating Experience. This oversight is not considered indicative of the licensee's current industry experience review program. The SIL stated that the condition could result in damage to the solenoid connector and recommended utilities rotate the solenoids to clear the interference. The SIL also indicated that the potential for the condition could be eliminated by tying off the solenoid cable to the HCU frame, thereby restricting rotation of the solenoid. VY initiated a work request (36773)to install tie wraps on the solenoid cables.

VY concluded that the interference condition would not have adversely affected the control rod scram time. The opening force of the scram valve is significantly greater than what would be required to bend the solenoid cable connection out of the way. Additionally, only a very small amount of stem travelis required to provide sufficient flow through the scram valve to achieve the required control rod scram time. Although the directional control valves do not perform a safety function, the potential for damage is of interest from an equipment reliability

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perspective and because the ability to insert individual control rods using the rod control system is an option in Emergency Operating Procedures for response to

"beyond design basis" events.

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Conclusions VY responded promptly to eliminate an NRC-identified mechanicalinterference between the solenoid for a directional control valve and the inlet scram valve on one hydraulic control unit (HCU). VY determined that the solenoid could have been damaged by operation of the scram valve but, the operability of the HCU was not affected. VY developed appropriate long term corrective action.

M8 Miscellaneous Maintenance issues M8.1 Review of Ooen items (92902)

The following open item was reviewed for closure based on a sampling of the licensee's corrective action _. -.

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(Closed) VIO 98-04-02: Procedure Non-Compliance During the 1938 refueling outage a procedure non-compliance resulted in a direct short being created across a main station battery, causing minor personnelinjuries and minor equipment damage. In response to this event, VY suspended work on energized electrical equipment and formed a task force to investigate the event.

The root cause of the event was determined to be personnel error, in that steps of

the continuous use procedure were performed simultaneously and therefore, out of sequence. As corrective action, VY management briefed workers on requirements for procedural adherence, and provided first line supervisors with additional leadership guidance material. The remaining outage work on energized electrical equipment received additional management oversight. The inspector concluded that

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these actions adequately addressed the causes of the event, and therefore, this violation is closed.

M8.2 In-office Review of LERs Related to Maintenance and Surveillance (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.

(Closed) LER 97-016-00: Missed Surveillance due to a Personnel Error as a Result of an Inadequate Procedure versus Technical Specifications Surveillance Evaluation Technical Specification 4.2, Surveillance Requirements, requires functional tests of protective instrument logic systems. Between August 28 and September 10,1997, the licensee determined that a number of contacts in safety related logic circuits in the following systems were not tested during logic circuit functional tests as required by TS: Primary Containment Isolation, Automatic Depressurization, High Pressure Coolant Injection, and Reactor Core Isolation Cooling. A logic circuit functional test, as defined in TS 1.0, requires the testing of all relays and contacts l

in a logic circuit from the sensor to the activated device. This discovery was made during a VY revi w in response to Generic Letter 96-01," Testing of Safety Related Logic Circuits." VY's testing program previously tested the logic circuit relays and i

their functions but, it did not specifically test the contacts. Tests performed as part of the immediate corrective actions verified the operability of the affected contacts.

The LER appropriately described the event, the root causes were clearly specified, and corrective actions adequately addressed this issue. The inspector determined that the failure to test certain individual relay contacts was not significant because the relays were tested, relay contacts are highly reliable, and the recent testing confirmed there was no impact on the relays' ability to perform their intended safety functions. However, the missed surveillances constitute a violation of NRC

requirements. This non-repetitive, licensee-identified, and corrected violation is being treated as a Non-cited Violation, consistent with S 3ction Vll.B.1 of the NRC

Enforcement Policy. (Closed) NCV 98-12-01: Missed Surveillance of Individual Relay Contacts (Closed) LER 97-022-00: Inadvertent Primary Containment Isolation System l

Actuation Due to a Spurious Spike on a Reactor Building Vent Radiation Monitor -

Repeat Event On November 15,1997, a spurious signal spike occurred on the "B" reactor building vent radiation monitor. As a result, the primary containment isolation system (PCIS) actuated causing the reactor building ventilation to isolate and an automatic start of the standby gas treatment system. Other systems that isolated as a result of the group 111 PCIS isolation signal were the containment vent and purge system, the containment air monitoring system, and the containment air dilution system. After radiation levels were checked and verified to be normal, the PCIS isolation was reset and the systems were returned to their normal alignments.

A similar event occurred earlier in 1997, as reported in LER 97-007. This earlier event was assessed by VY to be acceptable as an infrequent expected occurrence, and no long term corrective action was undertaken. In light of the November 1997 event, VY is evaluating additional actions. Any corrective actions will be in accordance with Maintenance Rule (10 CFR 50.65) requirements. Therefore, no additional NRC tracking of the issue is required. VY properly reported this engineered safety system actuation and no violation of NRC requirements was identified.

(Closed) LER 97-024-00,01.02: Inadequate Analysis and Guidance Allows the Failure of Vertical Support Columns in the Plant Cooling Tower Which Rendered the

Plant Alternate Cooling System Susceptible to a Seismic Induced Failure l

I While performing an inspection of the cooling towers, VY observed that two structural members in the Alternate Cooling System cell were degraded. As corrective action, VY identified the need for better repair guidance and improvements in the approach to cooling tower inspections. The degraded 4"x4"

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wooden posts were replaced. VY has scheduled a followup review to verify s

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corrective actions regarding the adequacy of inspection records, repairs, and procedures. The inspector concluded that VY appropriately entered this issue in the corrective action program and corrected the cooling tower deficiencies. The f

inspector noted that this discovery was captured, and is being tracked, in the

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licensee's Maintenance Rule Program. No further NRC action is required for this issue. No violations were identified.

(Closed) LER 98-016-00: Reactor Scram on High Water Level as a Result of a Stuck l

Open Feedwater Level Control Valve Due to a Cap Screw Lodged Underneath the l

Valve Disk l

An NRC Special Team was chartered to review this event and their findings were i

documented in Inspection Report 50-271/98-09, dated July 10,1998. The inspector found that the LER adequately described the event, its causes, and the l

corrective actions. No new information was identified based on a comparison of the LER and the Special Team's report. Based on the team's review of the event and VY's submittal of the required LER, no further inspection is warranted. This LER is administratively closed.

Ill. Engineering E1 Conduct of Engineering E1.1 (Opened) IFl 98-12-02: Potential for Post-LOCA Reactor Buildina Pressurization a.

Insoection Scope (37551)

Engineering products associated with the evaluation and resolution of licensee identified deficiencies were reviewed to assess the adequacy of corrective actions.

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Observations and Findinas The standby gas treatment system is part of the secondary containment system and is described in FSAR Chapter 5. FSAR section 5.3.4.2, " Safety Design Basis,"

states that the standby gas treatment system shall maintain a negative pressure in j

the Reactor Building so that any air leakage will be into the Reactor Building.

in early 1998, during evaluation of a qualification issue associated with the reactor

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building (RB) air lock doors, VY engineering questioned whether the post-LOCA heat up of secondary containment would result in pressurization of the RB. Engineers working on the RB door issue raised the issue because other plants have evaluated this phenomenon as part of secondary containment analyses. At VY, the original licensing basis off-site dose calculations assumed the RB was always at a negative pressure and that all post-accident releases would be elevated (stack release) and filtered through the standby gas treatment system. A preliminary evaluation of this new issue found that the RB would become slightly positive for a short period, but j

the evaluation of the peak RB pressure was not completed. A preliminary evaluation to bound the radiological consequences of this scenario was also performed. While this evaluation concluded that the offsite dose would be increased due to the resultant unfiltered ground-level release, the projected

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exposure was still well within 10 CFR 100 limits.

In May 1998, the VY licensing department concluded that pressurization of the RB was not part of the originallicensing basis of the plant and that no further evaluation was required. However, it was not clear whether the post-LOCA pressurization should be evaluated as a deficiency in the original analysis or whether this scenario was bounded by the original VY licensing basis. Pending further NRC review, this issue will be tracked as an inspector follow-up item (IFl 98-12-02).

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VY engineers iden'tified a scenario which could result in pnst-LOCA pressurization of.

l the secondary containment and this apparently was not evaluated in the original l

licensing basis. Preliminary VY evaluations indicate that pressurization would occur, L

but the analyses were not finalized because the VY licensing department concluded that secondary containment pressurization was not part of VY's original licensing basis. This issue requires further NRC evaluation and will be tracked as an inspector follow-up item.

E8 Miscellaneous Engineering issues E8.1 Review of Ooen items (92903)

(Closed) P2197-001: Automatic Valve Scram Solenoid Pilot Valve Failures On May 1,'1997, the Automatic Valves Corporation (AVCO) of Novi, MI, notified the NRC that a buzzing noise was indicative of metallic vibration within the scram solenoid pilot valve (SSPV) which could lead to degradation of the valve's performance..This issue is discussed in Section M2.2 of NRC inspection Report 50-271/97-04 and the report concluded that appropriate interim corrective actions had been taken by VY.

During the 1998 refueling outage, VY replaced all SSPV pilot assemblies with a new version that incorporated design improvements identified during AVCO's evaluation of an April 24,1997, test failure. None of the refurbished SSPV have exhibited the buzzing noise which was characteristic of the deficiency. Based on VY's final resolution of this issue, this item is closed.

(Closed) IFl 97-12-01i Post-LOCA Torus Temperature Post-LOCA torus temperature concerns were addressed in NRC Inspection Reports 50-271/98-06and 97-10. On April 14,1998, an escalated enforcement action and civil penalty were issued to VY (reference EA 97-531). On May 14,1998, VY provided a response to the escalated enforcement action and did not contest the violations. Final corrective action, in the form of a Technical Specification change request submittal, was submitted prior to the restart from the 1998 refueling outage. Based on the NRC actions taken, this IFl is administratively closed.

(Closed) IFl 97-12-02: Potential Condition Outside Design Basis The location of vacuum breakers for the turbine exhaust lines of the high pressure coolant injection (HPCl) and reactor core isolation cooling (RCIC) systems was j-questioned based on the potential for two types of water hammer events. This issue was inspected and documented in Section E2.1 of NRC Inspection Report 50-271/98-80. This IFl is administratively closed.

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(Closed) EA 97-531-02014: Failure to Address Residual Heat Removal (P.HR) Pump Motor Starting Restrictions (also reference eel 97-10-01 d)

RHR motor starting limits in VY procedures were based on a motor design ambient temperature of 86 F. In inspection report 50-271/97-201,the NRC identified that the licensee had failed to correctly translate the RHR corner room accident design temperature and RHR pump motor manufacturer's starting limits into operating procedure OP-2124, "RHR System." Inspection report 50-271/97-10 documented that the licensee initiated Event Report (ER) 97-0714, dated June 10,1997, to address this concern, because the RHR pump motors may be required to start in ambient temperatures as high as 148*F. The inspectors confirmed that a licensee's i

internal memorandum VYE 67/97, dated June 17,1997, indicated that the actual design temperature for the corner rooms was 155oF and the difference between the original GE design requirement of 148 F was insignificant. In their response letter BVY 98-73, dated May 14,1998, Vermont Yankee (VY) stated that the motor manufacturer was consulted in regards to the RHR and Core Spray pump motors starting capability. In addition, VY referred to the standard industry motor standard, NEMA Standard MG-1, " Motors and Generators," to justify two start attempts from the maximum design temperature and therefore reduced the number of recommended starting attempts from three to two.

The same response letter indicated that both the operating procedures and the surveillance procedures were revised to limit the number of motor starts. The inspectors confirmed operating procedures OP-2123, Rev. 29, and OP-2124, Rev. 46, and surveillance procedures OP-4123, Rev. 31, and OP-4124, Rev. 48, had been revised to adequately indicate the new motor starting limitation.s. This item was closed based upon the revised motor starting limitations.

(Closed) EA 97-531-03014: Failure of Nonsafety-Related Components on Emergency Diesel Generators (also reference eel 97-10-01 f)

The NRC identified that the licensee had failed to assure the capability of the non-safety-related control air equipment to support the operation of the safety-related diesel generators. in inspection report 50-271/97-10,the inspector verified that each diesel has its own pressure regulator. The licensee issued ER 97-0512, dated May 9.1997, to address this issue and replaced the questionable components on both emergency diesel units with dedicated safety-related components.

The inspectors reviewed the associated diesel generator jacket cooling water flow diagram and confirmed that the flow control valve diagram included the control air components, including the pressure regulator. The inspectors reviewed the safety class evaluation worksheets and the Maintenance Planning and Control (MPAC)

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database and confirmed the change in classification for the pressure regulators supporting the service water flow control valves for the diesel generators. This item is closed.

(Closed) EA 97-531-04014: RHR/ Low Pressure Coolant injection (LPCI) Flow Instrument Uncertainty (also reference eel 97-10-01 g)

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The NRC identified that the licensee had failed to include instrument uncertainty in the determination of RHR/LPCI flow test acceptance criteria. As documented in

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inspection report 50-271/97-10,the inspector confirmed that, in response to this concern, the licensee issued ER 97-0694, dated June 6,1997. Although the licensee found no operability concern, it did acknowledge that the uncertainties associated with the RHR flow required resolution. In their response letter, BVY 98-73, the licensee indicated they had revised the operating and surveillance procedure acceptance criteria to account for flow instrument uncertainty. The inspectors confirmed that OP-2123,2124,4123 and 4124 had been revised to account for instrument uncertainties. Th.s item was closed based on the revised acceptance criteria in the above revised procedures.

l (Closed) VIO 97-10-08: Loss of Station Battery Service Test Quality Records As documented in inspection report 50-271/97-10,the inspector confirmed that the licensee had lost the individual cell voltage (ICV) and the battery terminal voltage test records required by battery test procedure OP-4215 for the main station battery test performed on September 14,1996. During that inspeciton the inspector also confirmed that the special test printout, required to demonstrate battery operability, had also been lost.

In their response letter, BVY 98-33, dated March 6,1998, the licensee indicated this item was re-entered into their corrective action system as ER-98-0220," Loss of Quality Assurance (QA) Records," and committed to perform a self-assessment of the training provided to the administrative staff handling QA records. The licensee confirmed that the records cited in the violation, as well as other similar records for the B-ECCS-A and -B batteries during the same time period, were loat. The licensee's root cause evaluation to support ER-98-0220 confirmed that the records had not been processed in accordance with their administrative procedures.

The inspectors interviewed the work control supervisor and manager, who indicated that training had been presented to their administrative staff concerning attachments to completed work orders being identified as quality records.

The inspectors confirmed that the QA records for the latest surveillance tests performed on four safety-related batteries had been transferred from maintenance to

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document control, and confirmed that the computer data sheets were filed with the battery surveillance test work order records. Based on review of the above corrective actions, the item is closed.

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(Closed) EA 97-531-01033: Failure to issue a Licensee Event Report (LER) for Torus Temperature (also reference eel 97-10-09a)

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l Inspection report 50-271/97-201 documented that the licensee had issued an event notification (EN 30175) on March 26,1996, as required by 10 CFR 50.72(b)(1)(ii)(B), indicating that operation at suppression pool maximum operating temperature may have placed the unit outside the design basis. Amendment No. 88 to the Vermont Yankee Technical Specifications changed the maximum permitted

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that the licensee failed to follow-up the call to the NRC operations center with a i

submittal of a written licensee event report, as required by 10 CFR 50.73(a)(2)(vii),

because this event could have rendered both trains of a system required to remove l

residual heat inoperable.

l The inspectors confirmed that the licensee had issued LER 97-20 (including two

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supplements) and had benchmarked their reportability process against 11 other j

utilities which operated BWR plants. The licensee made minor editorial changes to l

procedure AP 0010," Notification and Reports Due," in Revision 27, dated March

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l 31,1998, to add guidance for the use of " Engineering Judgement" by reference to l

NUREG 1022, " Event Reporting Guidelines." The inspector concluded that this

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corrective action was reasonable and therefore, this item is closed.

(Closed) EA 97-531-10014: Failure to issue an LER for RHR Minimum Flow (also reference eel 97-10-09b)

The NRC discovered that the instruments required to establish RHR minimum flow did not include an allowance for the large instrument uncertainty in the minimum flow range. The licensee performed a reportability review but failed to recognize that the procedural inadequacy was reportable.

The inspectors confirmed that the licensee had issued LER 98-002 and had benchmarked their reportability process against 11 other utilities which operated BWR plants. The licensee made minor editorial changes to procedure AP-0010,

" Notification and Reports Due," in Revision 27, dated March 31,1998, to add guidance for the use of " Reasonable Assurance" by reference to NUREG 1022. The inspector concluded that this corrective action was reasonable and therefore, this item is closed.

(Closed) IFl 97-201-26: Instrument Drift in Instrument Uncertainty Calculations The NRC identified that the licensee had failed to revise instrument uncertainty calculations to include instrument drift factors. During inspection 50-271/97-10, l

the inspector confirmed that the licensee had previously started an instrument setpoint program, including an Instrument Drift Analysis Design Guide, dated

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l October 31,1996, but left this item open pending NRC review of the licensee's incorporation of the drift design guide into the setpoint design guide.

The inspectors confirmed that the licensee had developed an Instrument Uncertainty

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and Setpoints Design Guide, dated May 7,1997. Based on a cursory review of the

guidance, the inspector determined no additional review was warranted and t

therefore, this item is closed.

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(Closed) VIO 98-09-04: Lack of Overcurrent Relay Calibration Evaluations inspection report 50-271/98-09 documented that the overcurrent relays used in the protective circuits for safety-related motors were repeatedly found out-of-calibration and VY f ailed to initiate deficiency documents in their corrective action process.

tri the violation response letter, BVY 98-124, dated August 10,1998, VY

ocknowledged that there was a weakness in their trending program for electrical protective relays. The inspectors confirmed the licensee had entered this concern

into their corrective action program (ER 98-1451) and also confirmed the licensee's

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immediate corrective action consisting of a department memorandum on the subject, dated August 5,1998, which reviewed the concerns and provided interim guidance.

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The inspectors interviewed representatives from the Electrical and Controls Maintenance Engineering Department and observed work in progress to develop a computer based electrical protective relay calibration trending program. This program was being loaded with historical data for the 4 kV protective relays when the relay is due for a routine calibration. VY is also evaluating the need to include switchyard relays in this trending program. In addition, the licensee plans to revise procedure AP-0310, " Surveillance, Preventive and Corrective Maintenance Program," to further address the concerns of this item as long term corrective actions. Based on the licensee's immediate corrective actions and the developing long term corrective actions as part of their response to the ER, the inspectors l

concluded that the licensee was adequately addressing this concern and closed this item.

E8.2 In-office Review of LERs Related to Enaineerina (90712)

l An in-office review of the following licensee event reports (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.

(Closed) LER 97-006-03,04: Use of inadequate Design Specification /

Implementation Document During Initial Plant Construction Results in the Failure to Maintain Proper Electrical Separation of Electrical Cables LER 97-006 and Supplements 1 and 2 were previously reviewed and closed out in NRC Inspection Report 50-271/97-12. In Supplements 3 and 4, additional circuits which were not in conformance with cable separation requirements were listed and l

additionallong term corrective actions were added. The inspector found the additional long term corrective actions acceptable. Additionally, the inspector

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determined that the supplements did not change the significance of the issue or the previous NRC conclusion.

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(Closed) LER 97-020-00.01.02: An inadequate Technical Justification for a Proposed Change to Plant Technical Specifications Results in Operations Outside of the Bounds of Current Analyses Relative to Suppression Pool Temperature During LOCA.

The subject of the LER was first identified during NRC team inspection 50-271/97-201. NRC inspection 50-271/97-10dispositioned the team inspection findings and l

an escalated enforcement action and civil penalty of $55,000 were issued on April 14,1998, for violations surrounding the torus temperature issue. The licensee's

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l most recent evaluation of torus temperature related issues was reviewed in

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inspection report 50-271/98-06,and no problems were identified. Prior to restart

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from the 1998 refueling outage, VY submitted a TS change request to reduce the normal temperature limit from 100 F to 90 F, consistent with the current administrative controls. This TS change request was still under review at the close l

of this inspection period. The inspector determined that LER 97-020, and its supplements, were adequate. No further inspection is required for this LER because l

the corrective actions for the cited violations will be addressed individually, as open items.

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(Closed) LER 98-007-00.01: Group lli Primary Containment Isolation Signal

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l Potentially Blocked from Initiating an isolation of the Containment Air Monitoring i

isolation Valves due to a Potential Single Failure

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On March 18,1998, VY identified a potential single failure mechanism that would

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prevent the four containment air monitor (CAM) isolation valves from closing in l

response to a primary containment isolation system (PCIS) signal. Specifically, a

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single override switch had been included in the control circuitry during original construction, to provide the capability to reopen the four valves with a group lll PCIS isolation signal present; in this event, VY postulated that failure of the override switch could prevent the valves from closing in response to a PCIS isolation signal.

In response to this event, VY closed the four CAM isolation valves, thereby eliminating the need for the PCIS function. This rendered the CAM inoperable, and initiated a seven-day shutdown LCO per TS 3.6.C.2. The plant was shut down two i

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days later for commencement of the 1998 refueling outage. Corrective action to

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eliminate the switch from the circuitry, was completed prior to startup from the outage.

VY concluded that the cause of this event could not be determined due to the age (

of the event, but speculated that the postulated failure would not have been l

considered a credible event at the time of construction. The inspector concluded l

that VY appropriately questioned the original design and that their long term

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corrective action was appropriate. The inspector concluded that the design for the bypass switch (ie, whether it was ordinally considered as a potential single failure)

was unclear. However, VY took corrective actions commensurate with current industry practices. Because the switch was installed in accordance with the original design, no violation of NRC requirements was identified.

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E8.3 Review of Selected Event Reports a.

Inspection Scoce (92700)

The inspector conducted an onsite review of select 10 CFR 50.72 licensee event l

notifications (ens) and associated 10 CFR 50.73 licensee event reports (LERs).

l Several related VY event reports (ERs) were reviewed to assess the licensee's l

evaluation and corrective actions, including reportability determinations.

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Observations and Findinas -

Event Notification 31906 EN 31906, dated March 6,1997, described a condition that could compromise the residual heat removal (RHR) system operability and accident mitigation capability.

This event was evaluated by the licensee through its corrective action program by initiating plant event report ER 96-0241 on April 16,1996. The plant event report described two possible RHR system configurations that could result in operation outside the design bases as described in the UFSAR. The first configuration involved use of a branch connection of the RHR system to drain the torus. The licensee determined that this configuration had never been used at the plant and so did not involve operation outside the design bases. Administrative controls were i

I implemented to preclude this system configuration. The second possible configuration involved an operating procedure change that permitted use of an alternate keep-fill system for the RHR system that was not seismically qualified, nor did it include check valves or automatic isolation valves that would stop backflow from the RHR system, when in service, through the alternate keep-fill lines.

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This event was initially discovered by the licensee in April 1996 and reported to the I

NRC in LER 96-012 on May 16,1996. The NRC reviewed LER 96-012 previously in inspection reports (irs) 50-271/97-02and 97-10. The NRC concluded that the licensee's apparent root cause analysis and corrective actions were acceptable and appropriately described in the referenced LER. Further, based on the fact that this issue met the NRC Enforcement Policy for handling as a non-cited violation per Vll.B.3, "Old Design issue," NCV 97-10-11a was closed in IR 97-10.

During this inspection period, the plant event report was again reviewed to verify that the root causes were reasonable and that the LER accurately described the problem and the actions taken to correct the condition. Further, the inspector

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reviewed supplemental LER 96-012-01, submitted on August 6,1997. The inspector determined that the licensee's evaluation of this concern was appropriate.

During the review of this problem to determine the extent of condition, the licensee discovered a third example of RHR system configuration that would result in l

operation outside the design bases. This additional example was described in the

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supplement to the LER. The licensee attributed the root causes of this event to be:

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an inadequate initial design of the RHR / stem, which did not accommodate

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off normal operational line-ups, such as the need for alternate keep-fill lines;

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the lack of clear design bases information for the RHR system; and,

failure to adequately evaluate procedure revisions in regard to the design

bases of the system.

l In the initial LER, the licensee provided a preliminary root cause, (inadequate review of the design bases when preparing procedure changes) and initial corrective actions, including a change to the operating procedure to prevent using the i

alternate keep-fillline-up and implementation of a temporary modification to permit use of a new connection from the condensate transfer system to the RHR system

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using the normal keep-fill lines.

In the supplemental LER, the licensee provided the results of their root cause l

analysis, as described above, and provided additional corrective actions to address l

the weak design basis information and the potential system operating procedure reviews that may have similarly occurred. Except for the system design bases documentation reviews and validation, which will continue throughout calendar i

years 1998 and 1999, the licensee has completed the corrective actions for this j

event report. System engineering reviews of critical system operating procedures

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were completed and no additional problems were identified other than described in the supplemental LER. While some additionalinformation was provided in the supplemental LER, the general causes and corrective actions were the same as

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j previously evaluated by the NRC in inspection report 50-271/97-10. No additional enforcement action for this design problem is necessary. (Closed) LER 96-012-01:

Low Pressure Cooling Flow Could Po^entially Be Diverted Due to an inadequate Design Review Prior to Proceduralizing the Use of the System

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Event Notification 31949. and 32106:

These two event notifications resulted from the licensee's Individual Plant Examination of External Events (IPEEE). Event Notification 31949 was made on March 13,1997 and 32106 was made on April 8,1997. The events involved potential flooding vulnerabilities of switchgear rooms or ECCS corner rooms due to l

failures of non-seismic fire water system pipe during worst case seismic conditions.

l The licensee root cause analyses and corrective action evaluations were j

documented in plant event reports, ER 97-0250, and ER 97-0359. The events sulted in submittal of LER 50-271/97-04and two supplemental LERs providing additionalinformation about the extent of condition, completed root cause analysis j

and completion of additional corrective actions.

l In both cases the root cause of the events was the lack of a detailed flooding l

document describing the design and licensing requirements and the bases for those requirements. This inadequate bases resulted in modifications to the plant structures and systems that either created or changed the plant's vulnerability to internal flooding. The licensee's investigation concluded that the IPEEE analysis was of sufficient depth to question prior qualitative results and assumptions that led i

to the acceptance of the vulnerable design features.

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LERs 97-004-00 and 97-004-01 were previously reviewed and closed in NRC 1R 50-271/97-10,Section E8.3. In that report the NRC exercised discretion not to i

cite a violation of design control requirements because these were old design issues. An onsite review of supplemental LER 97-004-02 occurred during this current inspection. In this supplement, the event causes were reported and the long

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term corrective actions identified. The inspector found that the event causes were l

accurately identified and the corrective actions acceptable. Additionally, the j

inspector determined that the supplement did not change the significance of the

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l issue or the previous NRC conclusion. (Closed) LER 97-004-02: Lack of Detailed

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i Licensing Bases Document Relative to internal Flood Protection Combined with Errors in interpretation of Design Drawings Results in Configurations inconsistent

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with Plant Design Bases for Flooding Event Notification 34005 l

Event Notification 34005 was made on April 2,1998. This event involved a discovery by the licensee that the safety design basis for the ATWS mitigation

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system lacked specific supporting analysis. This 10 CFR 50.72 event notification l

was associated with plant event report ER 98-0680. Subsequent to identifying this (

concern, the licensee found sufficient supporting analyses conducted by their fuel l

vendor to determine that the safety design bases were accurate. The 10 CFR

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50.72 event notification was retracted and no additional corrective actions were taken.

The licensee discovered this problem during an ongoing engineering investigation of

the failure of the "B" reactor recirculating water pump field circuit breaker on March l

20,1998. Since this component is used in part of the ATWS mitigation system, the safety design bases for that function were reviewed as part of the licensee's

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investigation. While the safety design bases were clearly stated in both the UFSAR

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and in the design change package that installed the ATWS mitigation system, the l

design basis calculations that showed that these safety design bases could be met were not readily available or retrievable.

The Vermont Yankee system had been reviewed and approved by the NRC as a

"Monticello-type" system. The design basis calculations for Monticello were included in the design change package, but the Vermont Yankee specific calculations were not similarly available. On April 28,1998, the licensee completed its followup of this event and concluded that the Vermont Yankee design basis calculations had been previously completed by their vendor and were available to i

show that the accident mitigation system would prevent damage to the nuclear

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system process barrier as a result of an abnormal transient that does not result in a scram. Once the licensee located the design basis information and ensured that the safety design bases were accurate, the licensee retracted the 10 CFR 50.72 event notification.

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Event Notification 31923 Event Notification 31923 was made on March 10,1997. This event involved potential flooding of vital switchgear from leakage through open electrical conduit that could be submerged during worst case external flooding conditions. This 10 CFR 50.72 event notification was associated with plant event report ER 97-0197 and LER 50-271/97-002.

Plant event report ER 97-0197 described the apparent root cause as an inadequate initial design of the plant resulting in a vulnerability to external flood leakage into the j

switchgear room. In 1995 while performing a calculation oer the IPEEE for external

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flooding the licensee recognized that additional review w.s needed to ensure that external flooding did not seep through exposed conduit. When this design review was completed in 1997, the licensee determined that the vulnerability existed and reported that to the NRC. Additional discussion in the plant event report showed that the licensee had not considered externalleakage through conduit when following up on NRC Information Notices regarding potential flooding concerns.

Several missed opportunities to identify this problem earlier were described by the licensee. Overall, the licensee attributed this to an inadequate flood design basis.

The root cause was subsequently described in LER 50-271/97-002-01. Prior NRC inspections of this concern are described in NRC inspection report 50-271/97-03, Section E7.1, and inspection report 50-271/97-10,Section E8.3. In this latter report the NRC completed its review of LER 50-271/97-002 and 97-002-01, and concluded that the licensee had appropriately determined the root cause and taken corrective actions. A non-cited violation was issued for this problem m inspection report 50-271/97-10,as one of several examples of "old design issues." Based on review of the licensee's event report, no significant new information was identified that would change the prior conclusions about this event.

Event Notification 32016 and 33789

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Event Notification 32016 was made on March 25,1997, and 33789 was made on February 25,1998. The first event involved a potential overpressure failure of the standby gas treatment system if a LOCA would occur during containment purging operations due to the designed closure time of the associated containment isolation valves. This 10 CFR 50.72 event notification was associated with plant event report ER 97-0293 and LER 50-271/97-005. The second event involved a failure to maintain the corrective actions taken for the initial event. This 10 CFR 50.72 event notification was associated with plant event report ER 97-1040 and LER 50-271/97-014.

The licensee initially identified this concern on March 25,1997 as a result of a l

review of industry operational experience (an event at LaSalle on February 21, l

1997). Plant event report ER 97-0293 described that the apparent root cause of l

this event was an inadequate safety evaluation that was conducted to support a 1974 design change that removed interlock features that prevented purge and vent valves from being opened at power. A number of additional opportunities to discover this condition were missed, for example during procedure change reviews, but the licensee determined that the inability to retrieve the safety design bases for

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the affected systems and a lack of understanding of the pertinent licensing bases made it difficult for the staff to identify this concern.

The NRC reviewed and closed LER 97-005-00 and supplement 97-005-01 in inspection report 50-271/97-10. The NRC concluded that the root cause analysis and corrective actions were acceptable during that inspection. The NRC exercised discretion not to cite a violation for the "old design concern" in Section E8.3 of that report.

On April 24,1997, the licensee erroneously determined that the LaSalle event and design vulnerability of the standby gas treatment system was not applicable to Vermont Yankee based on a review of the design and licensing bases. As a result, administrative controls that were being used to prevent the vulnerable system configuration (inerting at power) were removed. The licensee then used that configuration with the unit at power on May 8,1997. On August 7,1997, while the licensee was completing the analysis that was to be used to retract event notification 32016,the licensee discovered additionallicensing bases information that clearly indicated that operation in this manner was unacceptable at Vermont Yankee.10 CFR 50.72 event notification 32735 was submitted to the NRC that same day describing this new event. LER 97-014-00 was submitted to the NRC on September 5,1997, which committed to an engineering review of the likely consequences to standby gas treatment system if a LOCA had occurred during inerting operations. Also, the NRC issued a Notice of Violation to Vermont Yankee in inspection report 50-271/97-06for f ailure to implement the corrective actions developed for the initial event that led to this subsequent event. The cause of this related event was further evidence of the misunderstanding of the licensing bases for this system.

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When the licensee completed the engineering analysis in February 1998,(a commitment in LER 97-014-00),it was determined that the standby gas treatment

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system filter housing design pressure could have been exceeded during inerting i

operations if a LOCA had occurred. An additional vulnerability was also discovered l

that could have had the same consequences for the routine use of the three-inch bypass lines used to maintain the drywell to torus differential pressure. The licensee concluded this meant that the plant had been operated inconsistent with the design bases and a 10 CFR 59.72 event notification was made on February l

25,1998 (Event Notification 33789). The licensee characterized this event as a lack l

of understanding of the plant licensing and design bases that resulted in inadequate i

response to industry operating experience which allowed plant operation inconsistent with the design bases.

LER 97-014-01 was issued on March 19,1998 for these related events in which the licensee concluded that the root causes and related causes were:

an inadequate disposition of the potential overpressure issue due to a lack of

understanding of the pertinent licensing and design bases.

the vulnerable design resulted from an inadequate analysis used to support a

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the pertinent licensing basis data / correspondence was not readily retrievable.

• The licensee's corrective actions included: establishing appropriate administrative controls for system configuration to preclude the alignments that could overpressurize the standby gas treatment system during accident conditions; completion of a full scope design vulnerability review for the standby gas treatment system for inerting operations at power coincident with a LOCA; providing training l

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and licensing bases; and providing better controls over the retrievability of licensing bases information.

Based on the review of the plant event reports, associated LERs and correspondence pertaining to the Notice of Violation issued with NRC inspection report 50-271/97-06,the inspector concluded that the licensee had provided reasonable root cause analysis and implemented appropriate corrective actions.

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Conclusion Review of these event reports indicated that the licensee had found several examples of original design vulnerabilities as a result of their IPEEE reviews. The licensee's root cause analysis has shown that these problems were either original design deficiencies or were a result of modifications to the plant that were based on evaluations where the design basis information was weak. Root cause analysis, corrective actions and reporting of these conditions were all found acceptable. As previously concluded in inspection report 50-271/97-10,the licensee's design basis documentation verification program should produce results similar to the Architect Engineering Team inspection and should be able to identify and correct similar conditions.

There were no current indications of weak operational experience feedback or safety evaluations. Licensee corrective actions for these events appropriately addressed prior concerns in these areas.

IV. Plant Support S8-Miscellaneous Security and Safeguards Issues S8.1 (Closed) VIO 98-05-01: Inadequate Package Search a.

Inspection Scope (81700)

To determine whether the conduct of security and safeguards activities met the licensee's commitments in the NRC-approved security plan (the Plan) and NRC regulatory requirements. The security program was inspected during the period of August 31-September 2,1998. Area inspected was protected area access control l

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Observations and Findinas

Protected Area (PA) Access Control of Packaaes: On September 1 and 2,1998,

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j the inspector observed the conduct of searches of packages at the access control

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point. All hand carried items ectering the PA were properly searched.

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j in response to violation 50-271/98-05-01,the licensee provided remedial training to l

the individual involved and self assessments were performed on the hand search

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practices of other officers. Lessons learned and reinforcement of performance expectations were communicated to each member of the security force. These corrective actions were reviewed and determined to be effective. The violation is

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closed.

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Conclusions l.

The licensee was conducting its security and safeguards activities in a manner that protected against acts of radiological sabotage and that this portion of the program, as implemented, met both the licensee's commitments and NRC requirements.

S8.2 (Closed) VIO 98-05-02: Intrusion Detection

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Inspection Scoce (81700)

Areas inspected were testing and maintenance of protected area detection aids, b.

Observations and Findinas PA Detection Aids: On September 2,1998, the inspector observed performance testing of the perimeter intrusion detection system (PIDS), at the protected area barrier (PAB). Testing consisted of fifteen attempts to climb the fence and tap testing (to simulate cutting) in eleven different zones. The testing, performed by the licensee and NRC program office personnel, resulted in the generation of appropriate intrusion alarms in all cases, in addition to the performance testing, the inspector reviewed corrective actions taken by the licensee, in response to violation 50-271/98-05-02 since the March 1998 inspection. All fence zone sensors analyzers were adjusted to detect climbing attempts. A specifically defined climb test has been incorporated into regularly i

scheduled operability testing of the PIDS. Finally, barbed tape has been installed to increase the delay afforded by portions of the perimeter fence. These corrective actions were determined to be effective. The violation is closed.

Testina and Maintenance: The inspector reviewed testing and maintenance records

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for the PIDS generated since the March inspection, and found the documentation

demonstrated that the licensee was testing and maintaining systems and equipment i

as committed to in the Plan.

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Conclusions The licensee's security facilities and equipment were detcimined to be well maintained and reliable and were able to meet the licensee's commitments and NRC requirements. Violation 50-271/98-05-02 is closed.

S8.3 (Closed) VIO 98-04-06: Potential Pathway for Unmonitored Vital Area Access During the 1998 refueling outage, the NRC identified a potential pathway for

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unmonitored access from the turbine building into the reactor building vital area, as discussed in inspection report 50-271/98-04. In response to this condition, VY implemented compensatory measures while a permanent physical barrier was constructed to completely enclose the access. The barrier was completed prior to plant restart. This violation is closed.

F2 Status of Fire Protection Facilities and Equipment F2.1 Fire Barrier Penetration Seals a.

inspection Scoce (90712)

An in-office review of LERs related to degradation of fire barrier penetration seals was performed to evaluate potential common causes and to assess VY's overall corrective actions. LERs98-001,98-008, and 98-014 were reviewed by the j

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inspector and discussed with the site Fire Protect.on Program Manager.

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Observations and Findinas in May 1997, VY initiated a fire barrier penetration seal upgrade project. The project encompassed the inspection of over 1800 penetration seals. This VY initiative has identified a number of non-conformances that, in some cases, indicate the subject seal would not act as a barrier for the intended duration (typically 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />). Because the fire barriers, and compensatory measures, are required by the VY Technical Specifications (TS), VY has made several 50.73 reports regarding past operation in a condition prohibited by TS. The inspector noted that some non-conformances appear to have been identified based on a better overall awareness of fire barriers than may have existed in the past.

In each case a fire barrier was degraded to some degree but, other fire protection features such as detection and suppression were available. Also, the on-site fire brigade is present twenty four hours a day to provide a prompt response.

(Closed) LER 98-001-00.01: A Technical Specification Fire Barrier Penetration Seal Was Determined to Be in a Condition Which Lowered the Rating of the Affected Fire Barrier VY personnel identified a fire penetration (grout block-out) surrounding a cable tray was not properly installed during its original construction. The fire barrier within the

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cable tray was acceptable. This original construction deficiency was appropriately documented in the corrective action program and the specific problem was corrected. VY has committed to long term corrective actions that will assess potentially similar instaliations that have not been disturbed since the original construction.

(Closed) LER 98-008-00.01: A TS Fire Barrier Penetration Seal Material Depth Did Not Meet the Requirements for a Three Hour Fire Barrier VY personnel identified that an penetration seal was not installed in accordance with its tested configuration. Although the proper depth of material existed, the seal was not fully contained within the confines of the wall. The depth of seal

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material within the confines of the wall was approximately half of the required depth and this invalidated the 3-hour rating of the barrier.

(Closed) LER 98-014-00,01: Inadequate Configuration Control Methods Result in Plant Operation With a Degraded Vital Fire Barrier Subject to Plant Technical Specification Requirements This LER reported VY's discovery that a seal repair, made sometime in the past, did not conform to the tested seal material configuration. The elastomer seal was repaired using a foam material. This issue was discussed in inspection report 50-271/98-08 and was dispositioned as a non-cited violation (NCV 98-08-09). This problem was attributed to weakness in the seal configuration control process that existed prior to the penetration seal upgrade project. The LER also reported that VY identified a second degraded seal where the physical configuration did not match the approved design. This problem was attributed to an installation drawing error regarding tne wall thickness.

Intearated Assessment The inspector noted that VY's on-going efforts to improve the fire barrier penetrations have been effective thus far. Corrective actions assigned to the LERs discussed above were considered reasonable, and the actions have been prioritized based on their relative importance. In each case described by VY, the penetrations would have retarded the propagation of a fire, but the as-found barrier would not likely have been able to withstand a 3-hour fire test. Based on the degradation, and l

defense in depth provided by the fire protection features, the inspector concluded l

the impact of these deficiencies, even when taken together, was minimal.

VY identified several fire barrier deficiencies which were not constructed to meet I

the approved 3-hour barrier configurations and the failure to have the required fire l

barriers constitutes a violation of Technical Specifications. A significant licensee

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initiative is currently in progress which is intended to enhance fire barrier penetration seals. Because the violations were identified as part of the licensee's

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on-going initiative, the individual deficiencies were not considered to be repetitive violations. This licensee-identified, non-repetitive, and corrected violation is being

treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (Closed) NCV 9812-03: Degraded 3-hour Fire Barriers c.

Conclusions VY initiatives identified several fire barrier penetrations in the plant that were not configured or constructed as designed. These findings were reported in Licensee Event Reports (LERs) and appropriate programmatic corrective actions are being taken. Although a fire barrier existed in each instance, their 3-hour rating was degraded. These deficiencies, individually or taken together, did not cause a significant degradation of the overall fire protection capabilities.

F8 Miscellaneous Fire Protections issues F8.1 In-office Review of LERs Related to Fire Protection (90712)

An in-office review of the following licensee event reports (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.

(Closed) LER 96-026-01: Inadequate Design Implementation and Subsequent Inadequate Documentation of Inspection Findings Result in Operation Outside of Plant Design Basis for Fire Mitigation and Technical Specification Non-compliance LER 96-026 was previously reviewed and closed in NRC Inspection Report 97-10.

In Supplement 1 to this LER, the licensee reported the completion of the corrective actions.

(Closed) LER 97-019-00: An inadequate Hydraulic Calculation Performed in Support of a 1982 Fire Protection Sprinkler System Modification Allowed Degradation of Redundant Emergency Diesel Generator Manual Sprinkler Subsystems An inadequate calculation was identified by VY during implementation of corrective actions for a previous event involving compliance with National Fire Protection Association code requirements. VY failed to recognize the licensing basis implications of a modification that reduced the capacity of diesel generator room

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sprinkler system. Appropriate compensatory measures were implemented after the issue was identified, however the condition existed for a long time before the

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compensatory fire watch required by Technical Specifications was implemented.

The inspector verified the Special Report required by TS 3.13.F.2 was completed (as part of the LER). Also, a sample of records for the compensatory firewatch rounds performed between September 24 and October 1 were reviewed against security access records.ind no discrepancies were identified.

Although the manual sprinkler system was inoperable, the system was still functional. VY appropriately reported this TS violation and has initiated a design

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l modification to restore the sprinkler system capacity. -The modification is scheduled

.for implementation by the end of 1998. This licensee-identified, non-repetitive, and..

corrected violation is being treated as a non-cited violation, consistent with Section l

Vll.B.1 of the NRC Enforcement Policy. (Closed) NCV 9812-04: Delay in Compensatory Firewatch implementation F8.2 Review of Open Items (92904)

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E (Closed) VIO 97-12-05: Automatic Self-Closing Fire Door Found inoperable -

This violation addressed an instance where the automatic self-closing feature of the j

high pressure coolant injection (HPCI) pump room fire door was discovered to be Inoperable. VY investigation determined that the most likely cause of this condition

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was that the door had inadvertently been pushed beyond its full open position; this lifted the closure mechanism drop weight above its normal position, which caused the cable restraining device to enter the associated pulley and walk the cable out of the pulley. The condition was corrected by installing a more streamlined cable

. restraining device. Based on no recurrences of problems with the door since completion of the modification, this violation is closed.-

V. Management Meetings X1 Exit Meeting Summary The resident inspectors met with licensee representatives periodically throughout the inspection and following the conclusion of the inspection on October 30,1998.

At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were discussed. NRC region based inspectors debriefed VY management at the conclusion of their respective on-site inspections. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

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NRC/ Vermont Yankee Public Meeting On September 16,1998, the NRC held a meeting with Vermont Yankee management regarding the results of the NRC's Systematic Assessment of Licensee

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. Performance (SALP) report. The SALP report was issued on August 28,1998.

This meeting was held at the Vernon, VT town hall and was open for public observation.

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Following the meeting with Vermont Yankee, NRC senior management remained at

- the town hall for discussion with members of the public and local media representatives.

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ITEMS OPENED, CLOSED, AND DISCUSSED i

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OPENED IFl 98-12-02:

Potential for Post-LOCA Reactor Building Pressurization (page 9)

l CLOSED VIO 98-09-01:

Plant Operating Procedures (page 3)

LER 97-023-00:

A Component Failure in the Main Generator Protection Circuitry Results in a Reactor Scram (page 3)

VIO 98-04-02:

Procedure Non-Compliance (page 7)

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LER 97-016-00:

Missed Surveillance due to a Personnel Error as a Result of an inadequate Procedure versus Technical Specifications

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Surveillance Evaluation (page 7)

LER 97-022-00:

Inadvertent Primary Containment isolation System Actuation Due to a Spurious Spike on a Reactor Building Vent Radiation Monitor - Repeat Event (page 8)

LER 97-024-00,01,02:

Inadequate Analysis and Guidance Allows the Failure of Vertical Support Columns in the Plant Cooling Tower Which Rendered the Plant Alternate Cooling System Susceptible to a Seismic Induced Failure (page 8)

LER 98-016-00:

Reactor Scram on High Water Level as a Result of a Stuck Open Feedwater Level Control Valve Due to a Cap Screw Lodged Underneath the Valve Disk (page 8)

P2197-001:

Automatic Valve Scram Solenoid Pilot Valve Failures (page 10)

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IFl 97-12-01:

Post-LOCA Torus Temperature (page 10)

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IFl 97-12-02:

Potential Condition Outside Design Basis (page 10)

l EA 97-531-02014:

Failure to Address Residual Heat Removal (RHR) Pump Motor

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Starting Restrictions (also reference eel 91-10-01d)(page 11)

l EA 97-531-03014:

Failure of Nonsafety-Related Components on Emergency ~ Diesel

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Generators (also reference eel 97-10-01f)(page 11)

EA 97-531-04014:

RHR/ Low Pressure Coolant injection (LPCI) Flow Instrument Uncertainty (also reference eel 91-10-01g)(page 12)

VIO 97-10-08:

Loss of Station Battery Service Test Quality Records (page 12)

EA 97-531-01033:

Failure to issue a Licensee Event Report (LER) for Torus Temperature (also reference eel 9710-09a)(page 12)

EA 97-531-10014:

Failure to issue an LER for RHR Minimum Flow (also reference

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eel 97-10-09b)(page 13)

IFl 97-201-26:

Instrument Drift in Instrument Uncertainty Calculations (page 13)

VIO 98-09-04:

Lack of Overcurrent Relay Calibration Evaluations (page 14)

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LER 97-006-03,04:

Use of inadequate Design Specification / Implementation l

Document During Initial Plant Construction Results in the Failure to Maintain Proper Electrical Separation of Electrical Cables (page 14)

LER 97-020-00,01,02:

An inadequate Technical Justification for a Proposed Change to Plant Technical Specifications Results in Operations Outside

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Pool Temperature During LOCA (page 15)

LER 98-007-00,01:

Group lll Primary Containment isolation Signal Potentially Blocked from Initiating an isolation of the Containment Air l

Monitoring Isolation Valves due to a Potential Single Failure

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(page 15)

LER 96-012-01:

Low Pressure Cooling Flow Could Potentially Be Diverted Due l

to an inadequate Design Review Prior to Proceduralizing the Use of the System (page 17)

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l LER 97-004-02:

Lack of Detailed Licensing Bases Document Relative to Internal Flood Protection Combined with Errors in Interpretation of Design Drawings Results in Configurations inconsistent with Plant Design Bases for Flooding (page 18)

VIO 98-05-01:

Inadequate Package Search (page 21)

VIO 98-05-02:

Intrusion Detection (page 22)

l VIO 98-04-06:

Potential Pathway for Unmonitored Vital Area Access (page 23)

LER 98-001-00,01:

A Technical Specification Fire Barrier Penetration Seal Was Determined to Be in a Condition Which Lowered the Rating of l

the Affected Fire Barrier (page 24)

LER 98-008-00,01:

A TS Fire Barrier Penetration Seal Material Depth Did Not Meet the Requirements for a Three Hour Fire Barrier (page 24)

LER 98-014-00,01:

Inadequate Configuration Control Methods Result In Plant i

Operation With a Degraded Vital Fire Barrier Subject to Plant Technical Specification Requirements (page 24)

LER 96-026-01:

Inadequate Design implementation and Subsequent inadequate Documentation of Inspection Findings Result in Operation Outside of Plant Design Basis for Fire Mitigation and Technical Specification Non-compliance (page 25)

LER 97-019-00:

An inadequate Hydraulic Calculation Performed in Support of a 1982 Fire Protection Sprinkler System Modification Allowed Degradation of Redundant Emergency Diesel Generator Manual Sprinkler Subsystems (page 25)

VIO 97-12-05:

Automatic Self-Closing Fire Door Found Inoperable (page 26)

NON-CITED VIOLATIONS OPENED / CLOSED NCV 98-12-01:

Missed Surveillance of Individual Relay Contacts (page 8)

NCV 98-12-03:

Degraded 3-hour Fire Barriers (page 25)

NCV 98-12-04:

Delay in Compensatory Firewatch Implementation (page 26)

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LIST OF ACRONYMS USED A/E Architect / Engineering BMO Basis for Maintaining Operation BWR Boiling Water Reactor CAM Containment Air Monitor CFR Code of Federal Regulation CS_

Core Spray EDG Emergency Diesel Generator EN Event Notification EPR Electronir Pressure Regulator ER Event Report FSAR Final Safety Analysis Report GE General Electric HCU Hydraulic Control Unit HPCI High Pressure Coolant injection IFl Inspector Follow Item JPM Job Performance Measure LCO Limiting Condition for Operation LER Licensee Event Report LOCA Loss of Coolant Accident LPCI Low Pressure Coolant injection LORT Licensed Operator Requalification Training NCV Non-Cited Violation NRC Nuclear Regulatory Commission PA Protected Area PCIS Primary Containment isolation System PIDS Perimeter intrusion Detection System QA Quality Assurance RB Reactor Building RHR Residual Heat Removal SALP Systematic Assessment of Licensee Performance SGTS Standby Gas Treatment System SIL Service Information Letter SSPV Scram Solenoid Pilot Valve TS Technical Specifications URI Unresolved item VIO Violation VY Vermont Yankee