IR 05000271/1997012
| ML20216B023 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 03/03/1998 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20216A971 | List: |
| References | |
| 50-271-97-12, NUDOCS 9803120395 | |
| Download: ML20216B023 (30) | |
Text
{{#Wiki_filter:, . U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.
50-271 Licensee No.
DPR-28 Repon No.
97-12 Licensee: Vermont Yankee Nuclear Power Corporation Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: December 7,1997 - January 24,1998 Inspectors: William A. Cook, Senior Resident inspector Edward C. Knutson, Resident inspector i Approved by: Curtis J. Cowgill, Ill, Chief, Projects Branch 5 Division of Reactor Projects I . l
1 i (DR803120395 980303 g ADOCK 05000271 PDR _ _ _ _ j
. . EXECUTIVE SUMM? RY Vermont Yankee Nuclear Power Station NRC inspection Report 50-271/97-12 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a seven week period of resident inspection.
Ooerations The cold weather protection program at VY establishes measures that are generally adequate to protect safety-related systems against extreme cold weather. The addition this year of a monthly verification checklist was a significant program enhancement.
Maintenance Special tests were conducted to collect information on operation of the RHR and CS pumps in the current minimum flow configurations, in response to concerns raised by the NRC in inspection report 50-271/97-201. The inspector observed that the tests were well planned and performed in a controlled manner. Due to a problem with the flow measuring device during the CS test, systems engineers attempted to obtain pump minimum flow line flow data during a subsequent CS surveillance. This non-intrusive data collection / ultrasonic instrument troubleshooting activity was appropriately controlled via a station work order.
During routine rounds, an auxiliary operator heard a noise in the feedwater heater bay which was subsequently determined to be an instrument air leak between the regulator and the valve controller for a feedwater heater high level dump valve. A repair plan and operational contingencies were promptly developed, and the repair was completed without incident. The inspector concluded that the air leak repair activity was an appropriately prompt response to a problem that had been identified as a result of good watchstanding practice.
An intermittent low electrical ground on 4160 volt AC bus 5 developed after startup from the November.1997 forced outage. Although bus 5 is not a safety class electrical power source, a ground fault is a :oncern because it could result in loss cf the associated startup transformer, which is the normal power supply to emergency bs 4. Troubleshooting corifirmed that the low ground condition was intermittent and of short duration. VY is examining strategies to identify and repair or eliminate the source of the low ground, including the possibility of conducting these activities while on-line.
Enaineerina During an independent review of the containment re-analysis and the torus temperature analysis of record, the licensee identified several potential sources of energy input to the torus that had either not been accounted for or were.on-conservative. Resolution of this issue will be tracked via an inspector follow-up item (IFl 97-12-01).
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__ . . . As a result of the design basis documentation effort, VY identified that the vent side of the vacuum breakers on the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam exhaust lines to the torus should more appropriately communicate from inside the torus vice outside primary containment. VY's RCIC and HPCI exhaust line vacuum breaker placement is different than most boiling water reactors, which have the valves located inside the torus. The VY configuration was the result of a 1973 design change to address unwarranted vacuum breaker operation during integrated leak rate testing. An inspector follow-up item (IFl 97-12-02)will track the licensee's final resolution of this issue.
Based upon the inspector's review of active Basis for Maintaining Operations (BMOs) and the VY staff's current projections for closure of BMOs following the Spring 1998 refuel outage, the NRC was concerned with resolution of the issues.
in reviewing seismic design considerations for the i24 volt DC power system, the inspector determined that there is no regulatory requirement for this system to be a safety class electrical system, and therefore, that the batteries need not be seismically mouWd.
Imprecise wording and inconsistencies in formatting in the FSAR which led to this issue being raised are being addressed by the licensee. Accordingly, unresolved item URI 97-05-03 is closed.
Seismic and safety grade classification issues associated with the HPCI gland exhaaster discharge line to the SBGT system, the torus /drywell pumpback system, and interfacing connections, were being appropriately addressed. Unresolved item URI 97-05-04 was closed based on the licensee providing reasonable basis that the classification was appropriate.
VY's identification and corrective actions to address electrical cable separation "old design issues" were appropriate and were executed, or planned to be accomplished, in a time period commensurate with their safety significance. These actions are consistent with Generic Letter 91-18 and 10 CFR 50, Appendix B, Criterion XVI, " Corrective Actions."
The extended period without appropriate electrical cabling separation was a violation of 10 CFR 50, Appendix B, Criterion Ill, Design Control, as well as the VY UFSAR. However, this violation was not cited in accordance with Section Vll.B.1 of the NRC Enforcement Polic r. (NCV 9712-04) Plant Sucoort The HPCI room automatic self-closing fire door was inoperable for an indeterminate period between October 2,1997 and January 13,1998. VY took prompt action to restore operability and to correct the suspected cause. The inoperable fire barrier was a violation of 10 CFR 50 Appendix R requirements. (VIO 97-12-05) iii
-. . TABLE OF CONTENTS EX EC UTIVE SU MM ARY.............................................. ii Summary of Plant Status............................................ 1 1. Operations ....................................................1
Operational Status of Facilities and Equipment................... 1 02.1 Cold Weather Preparations............................. 1 II. M aintenance................................................... 3 M1 Conduct of Maintenance.................................. 3 M1.1 Maintenance Observations........ ..................3 M1.2 Surveillance Observations............................. 4 M2 Maintenance and Material Condition of Facilities and Equipment....... 5 ~~ M2.1 Intermittent Low Electrical Ground on Bus 5................ 5 M8 Miscellaneous Maintenance issues........................... 6 M8.1 (Closed) Unresolved item 97-02-09: Instrumentation loop analytical limits exceed the Technical Specifications (TS) allowable value.... 6 Ill. Eng inee ring................................................... 6 E1 Conduct of Engineering.................................... 6 E1.1 (Open) Inspector Follow-Up item 97-12-02: Post-LOCA torus temperature determined to De higher than previously analyzed... 6 E1.2 (Open) Inspector Follow-Up item 97-12-03: Potential condition outside design basis involving location of vacuum breakers for the HPCI and RCIC steam exhaust lines........................ 7 E2 Engineering Support of Facilities and Equipment..................9 E.2.1 Heview of Basis for Maintaining Operations (BMOs) Proposed by VY ! to Remain Open Following the 1998 Refuel Outage........... 9 E8 Miscellaneous Engineering issues............................ 15 E8.1 (Closed) Unresolved item 97-05-03: Seismic design considerations for the 24 volt DC power system...................... 15 E8.2 (Closed) Unresolved item 97-05-04: Interface of a retired-in-place ' system with the high pressure coolant injection system....... 16 E8.3 (Closed) Unresolved item 97-03-02: Electrical cable separation issue ..............................................17 IV. PI' ant Support ................................................19 F2 Status of Fire Protection Facilities and Equipment................ 19 F2.1 Automatic Self-Closing Fire Door Found inoperable.......... 19 V. Management Meetings.......................................... 20 X1 Exit Meeting Summary................................... 20 X3 Review of Updated Final Safety Analysis Report (UFSAR)........... 20 INSPECTION PROCEDURES USED..................................... 21 iv
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I - ITEMS OPENED, CLOSED, AND DISCUSSED.............................. 22 i PARTIAL LIST OF PERSONS CONTACTED............................... 23 I LIST OF ACRONYMS USED.......................................... 24 ) V
. . Report Details Summarv of Plant Status At the beginning of the inspection period, Vermont Yankee (VY) was operating at 100 percent power. On December 16 and 30, and January 13, power was reduced to 70 to 80 percent for control rod pattern adjustments. With the exception of power reductions to conduct planned surveillance testing, the plant operated at 100 percent power until January 16, when the plant commenced a gradual power reduction due to fuel depletion (coastdown). The 1998 refueling outage is scheduled to commence on March 21.
During this inspection period, the inspectors reviewed the INPO assessment report, dated September 1997 which summarized INPO's two week assessment completed in July I 1997. The report did not contain VY's response. This documented assessment of.
licensee performance was consistent with the preliminary findings, as discussed with the i inspectors on August 8,1997 by the Plant Manager. The inspectors noted that the INPO { findings were also generally consistent with the most recent NRC staff assessments of the licensee's performance.
1. Operations
Operational Status of Facilities and Equipment' 02.1 Cold Weather Preparations a.
jmgection Scoce (71714) i i The inspector reviewed the licensee's program to protect safety-related systems i against extreme cold weather.
i b.
Observations and Findinos The governing procedure for VY's cold weather protection program is OP-2196, " Preparations for Cold Weather Operations." The procedure is to be initiated when average ambient temperatures are less than 35 degrees F, weather forecasts indicate the onset of cold weather, or during the week of October 15. Required preparations for the operations, maintenance, and instrument and controls (l&C) departments are specified in attachment 1 to OP-2196, " Cold Weather Initiation Operations Checklist." Following completion of this checklist (nominally during the first two weeks in November), the operations department performs a verification of cold weather preparations using attachment 3 to OP-2196," Operations Cold Weather Protection Verification Checklist." Subsequent verification that cold weather protection measures are in place and operating properly is performed by operators during normal watchstanding tours. Normal operations are restored when weather conditions dictate.
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic l . .
The inspector had previously reviewed VY's cold weather protection program for the 1996/1997 winter season, as discussed in inspection report 50-271/96-11.
During the current inspection, the inspector noted that two revisions had been made to OP-2196 since the previous inspection. The first revision (revision 10 and interim change DI 97-196) dealt with individual systems (auxiliary chemical j treatment system, service water bay de-icers, and service air system) in response to l VY event reports concerning cold weather problems with those systems. This was the effective revision at the time that the initiation and verification checklists were performed for the current winter season. Subsequently, revision 11 was incorporated. Among other things, this revision added a requirement for monthly verification of cold weather protection measures in accordance with an additional checklist. The inspector considered this to be a significant program enhancement.
The inspector reviewed the completed " Cold Weather Initiation Operations i Checklist" (attachment 1 to OP-2196) and " Operations Cold Weather Protection Verification Checklist" (attachment 3 to OP-2196) for 1997. The operations checklist was initiated on October 24 and completed on December 15; the verification checklist was initiated on December 23 and completed three days later.
Four work order requests (WORs) were initiated as a result of material deficiencies identified while performing these checklists. The inspector reviewed the status of these WORs. Three had been completed (repairs to two steam heating coils in the condensate storage tank pipe trench and a heater in the chemical treatment shed).
The one outstanding WOR was for a defective fire hydrant drain valve, which will require excavation of the hydrant for replacement. From discussion with maintenance planning personnel, the hydrant remains operable, but would require manual draining, were it to be used, to return it to the standby condition.
The inspector performed a partial walkdown of plant buildings and outside support equipment to assess the extent and effectiveness of cold weather preparations.
The inspector noted that heating and ventilation in the intake structure was not coordinated. Specifically, in the circulating water (CW) pump room, the thermostats for the roof ventilator, which also operates the room east side louvers and the room heater are both located near the motor vent for the "C" CW pump. This pump was operating at the time of the inspection, and the warm air blowing on the thermostats resulted in the heater remaining continuously shut down, while the ventilator cycled. In effect, room heating was being provided solely by the operating equipment, and the roof ventilator was running unnecessarily, causing room temperature to be lower than desired. In the service water pump room, both the room heater and the roof ventilator were cycling. The inspector discussed these observations with the operations manager. The licensee acknowledged the inspectors observations. The inspector observed r'o other weather-related issues.
c.
Conclysigns The cold weather protection program at VY establishes measures that are adequate to protect safety-related systems against extreme cold weather. The add; tion this year of a monthly verification checklist was a significant program enhancement.
The implementing checklists for this program were initiated and completed later in
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the season than was the intent of the implementing procedure; however. the inspector observed no weather related problems as a result of the delay.
11. Maintenance M1 Conduct of Maintenance ' M 1.1 NIaintenance Observations a.
In_soection Scoce (62707) The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utili7ed, 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.
b.
Observations, Findinas, and Conclusions The inspector observed all or portions of the following maintenance activities:
i Residual heat removal (RHR) oumo minimum flow stecial test, observed
December 9.
Core scrav (CS) oumo minimum flow special test, observed December 19.
- The minimum flow tests were conducted to collect information on operation of the RHR and CS pumps in the current minimum flow configurations, in response to concerns raised by the NRC in inspection report 50-271/97-201.
The inspector noted that the pre-test briefs were well conducted, and that completion points as well as abort contingencies were clearly established.
The inspector observed no problems in the conduct of the tests, although flow data was not obtained during the CS test due to technical difficulties with the ultrasonic flow measuring equipment. The inspector concluded that the tests were well planned and performed in a controlled manner.
- No.1 feedwater heater dumo valve controller air leak reoair, observed pecember 31.
During routine rounds on December 31, an auxiliary operator heard a noise in the feedwater heater bay (a locked high radiation area, not immediately accessible), in a subsequent entry into the area, the source was determined to be an instrument air leak between the regulator and the valve controller for the no.1 feedwater heater high level dump valve. A repair plan was promptly developed which used a dummy air signal to the controller to ) maintain the dump valve closed while the repair was performed. The licensee recognized that an unanticipated problem during the repair could
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result in a reduction or loss of feedwater heating, which could produce a significant reactor plant transient. Actions for a loss of feedwater heating were reviewed by the control room operators prior to the repair, and a pre- , job brief was conducted in the control room. Reactor power was reduced by ' two percent to avoid exceeding rated power if a loss of feedwater heating were to occur.
The inspector observed that operators closely monitored plant parameters during the maintenance. No problems were experienced and the activity had no effect on plant stability. The inspector concluded that the air leak repair activity was an appropriately prompt response to a problem that had been identified as a result of good watchstanding practice.
M1.2 Surveillance Observations a.
Insoection Scoce (61726) The inspuctor 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, b.
Observations. Findinos. and Conclusions The inspectors observed portions of the following surveillance testing activities: Core sorav system surveillance. observed Januarv 6.
- The inspector observed various aspects of the "A" and "B" core spray (CS)
pump quarterly surveillance test conducted in accordance with Operating Procedure (OP)-4123, " Core Spray System Surveillance" revision 30, dated August 22,1997. The inspector witnessed the pre-evolutionary briefing held
in the control room and observed that the briefing was thorough. All l involved parties were appropriately briefed on their individual responsibilities and provided clear instructions, including those in the event of a testing problem. The senior control room operator clearly and concisely outlined the testing activities and established the communication and control functions.
i The inspector observed operation of both CS trains including pump vibration monitoring and lubricating oil sampling following pump shutdown. During , testing of the "B" CS train, systems engineers attempted to obtain pump minirnum flow line flow data using a strap-on ultrasonic measuring device.
! This non-intrusive, troubleshooting activity was appropriately controlled via a station work order. The inspector discussed various aspects of the surveillance test with operators, mechanics, and systems engineers and concluded that these individuals were knowledgeable of their testing responsibilities and that they were properly conducting the test per OP-4123.
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Reactor core isolation coolina system surveillance testina, observed Januarv
IL The inspectors observed various aspects of the quarterly reactor core isolation cooling (RCIC) system surveillance test conducted in accordance with OP-4121, " Reactor Core Isclation Cooling System Surveillance," j revision 36, dated July 17,1997. The inspectors observed that the pre-
evolutionary briefing was well attended and well conducted. The inspectors ' witnessed a part of the test from the RCIC room and noted that participants were on-station and prepared in advance of the actual pump start. The inspector also noted good radiation protection technician monitoring and control of testing participants and observers, in order to maintain personnel exposure ALARA due to the RCIC steam supply line dose rates during system operation. The inspectors concluded that overall conduct of the RCIC quarterly surveillance test was good.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Intermittent Low Electrical Ground on Bus 5 a.
Inspection Scope (92902) l l The inspector reviewed VY's actions for an intermittent low electrical ground that I developed on bus 5.
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Observations and Findinas i
Bus 5 is part of the 4160 voit AC on-site distribution system at VY. The bus receives power from one of the two startup transformers (T-3B) while the plant is f shut down, or from the main generator through unit auxiliary transformer T-2 during ! plant operations at power. Bus 5 is not a safety class electrical power source.
! However, a ground fault on bus 5 is a concern because it could result in loss of the startup transformer, which is the normal power supply to emergency bus 4. A single low ground on the 4160 volt AC system will not necessarily result in a fault ! to ground, but is indicative of a degraded condition such as insulation breakdown.
Indications of the low electrical ground on bus 5 were first observed in November 1997, following startup from a forced outage. Typically, the condition existed just long enough to cause the main control board annunciator to alarm, and the alarm would clear as soon as it was acknowledged. Troubleshooting with test equipment confirmed that the low ground condition was intermittent and of short duration. At the close of the inspection period, VY was examining strategies to identify and repair or eliminate the source of the low ground, including the possibility of conducting these activities while on-line.
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Conclusions ( The inspectors will examine VY's plan for additional troubleshooting during the next
inspection period.
M8 Miscellaneous Maintenance issues M8.1 (Closed) Unresolved item 97-02-09: Instrumentation loon analvtical limits exceed the Technical Soecifications (TS) allowable value As documented in section M8.4 of inspection report 50-271/97-02,this issue was left unresolved pending the NRC staff's review of the completion of the licensee's Instrument Set point Program (ISP), being conducted in conjunction with the Improved Technical Specification (ITS) efforta. The current instrument setpoint ! values were demonstrated by the licensee to be within VY's licensing basis.
l Therefore, this matter was found acceptable. Any subsequent revision of these instrument setpoints and " analytical limits," as determined by VY's ISP and ITS programs, will be subject to NRC staff review and approval as established by the VY operating license. Accordingly, URI 97-02-09 is closed.
111. Engineering E1 Conduct of Engineering E1.1 (Onen) Insoector Follow-Up Item 97-12-02: Post-LOCA torus temperature determined to be hiaher than oreviousiv analyzed a.
insoection Scoce (37551,71707) On January 7, at 3:10 p.m. the control room staff made a 10 CFR 50.72 notification involving the preliminary analytical determination that the peak torus temperature, post-LOCA, could potentially exceed the containment design temperature limit of 176 degrees F. Th3 inspectors conducted a follow-up review of VY's actions. The inspectors examined the licensee's initial operability determination, as documented in Event Report (ER) No. 98-0025, dated January 6, 1998, and reviewed and discussed the summary analysis of the issue with the responsible ee.gineers, b.
Observations and Findinas During an independent review of the containment re-analysis and the torus temperature analysis the licensee identified that the computer code FROSSTEY-2, used to determine input to VYC-1290, contained several errors. The cumulative affect of these analytical and computer program errors was that the original post-LOCA maximum torus temperature unit of 176 degrees F may be exceeded by approximately 7 degrees F (183 degrees F) under certain conditions. Additionally, the short-term (10 minutes) post-LOCA suppression pool temperature may exceed
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the original temperature value of 155 degrees F by approximately 5 degrees F (160 degrees F).
The inspector reviewed VY's torus (primary containment) operability determination which concluded that, as long as indicated torus water temperature remained s80 degrees F and river water temperature remained s50 degrees F, the postulated post-LOCA primary containment temperatures would remain within the analyzed { limits. These more restrictive administrative limits (Technical Specification torus j water temperature limit of $ 100 degrees F and analytical river water temperature
limit of s 85 degrees F) imposed by the licensee were also reviewed with the l Region I technical staff on January 8 via a telephone conference call. These new administrative limits were determined to be appropriate interim actions pending a final resolution by the licensee's engineering staff.
At the time of VY's discovery of their containment analysis heat input errors, the , VY staff was preparing a licensing submittal for a revised Technical Specification torus temperature limit and their 90-day response to NRC Generic Letter 97-04, " Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pump," dated October 7,1997. These two submittals i had been postponed at the conclusion of the inspection.
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Conclusions The licensee's actions to identify, report, and implement more restrictive administrative operating limits on the torus temperature and river water tsmperature due to the identified containment analysis errors were found to be appropriate by the Region i technical staff. Continued safe plant operation with the 50 degree F river water temperature administrative limits provides an appropriately conservative operating margin for the containment heat removal capacity of the RHR heat exchangers. Licensee resolution of this issue will be tracked via an inspector follow-up item (IFI 97-12-02).
E1.2 (Ocen) Inspector Follow-Up Item 97-12-03: Potential condition outside desion basis involvino location of vacuum breakers for the HPCI and RCIC steam exhaust lines a.
Backoround and insoection Scope (92903,37551) On January 15, the VY engineering staff completed a preliminary review which concluded that the vent side of the 3-inch vacuum breakers on the high pressure coolant injection (HPCI) 24-inch steam exhaust line and reactor core isolation cooling (RCIC) 8-inch steam exhaust line to the torus should more appropriately commursicate from inside the torus vice outside primary containment. The basis for this preliminary conclusion was that, at elevated torus pressures (as would potentially exist during LOCA accident scenarios) with the vacuum breakars venting from the reactor building (secondary containment), potentially more water would be drawn into the HPCI and RCIC exhaust lines following these systems being secured, resulting in potentially more severe water hammer transients. The severity of the water hammer transients may challenge further HPCI and RCIC system operation
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and/or containment integrity. Based upon this preliminary engineering evaluation, the VY staff initiated an event report (No. 98-0057) and reported the condition to NRC as required.
J b.
Observations and Findinas The itsue was identified during the licensee's design basis documentation review effort. The prime contractor for this effort raised a question that VY's RCIC and HPCI exhaust line vacuum breaker placement is different than most boiling water reactors (BWR). As discussed above, the potential consequences of this design configuration could involve severe water hammer transients on the associated steam piping and supports. The VY engineering staff review identified that the vacuum breakers for both RCIC and HPCI had been relocated via Engineering Design Change Report (EDCR) No. 73-32 from inside the torus to outside the torus. The inspectors determined that (per EDCR 73-32) the basis for relocating the vacuum breakers in 1973 was to address the failure of the integrated leak rate test due to apparent unwarranted vacuum breaker operation during the testing. Engineering stated that review of EDCR No. 73-32 identified no discussion of the impact on vacuum breaker operation due to elevated torus pressures.
The inspectors examined the licensee's initial HPCI and RCIC operability determination, based upon this identified vacuum breaker concern, and concluded that it was acceptable. The inspectors noted that the operability basis was principally founded upon the fact that historical records demonstrated that, during pre-operational testing, both RCIC and HPCI systems experienced water hammer transients without damage to structural supports or piping. This pre-operational testing was performed prior to VY, or any other BWR licensee, having installed vacuum breakers on the HPCI or RCIC exhaust lines. The VY engineering staff referenced EPRI NP-6766, Volume 5, Part 1, Water Hammer Prevention, Mitigation, and Accommodation,"in deriving this conclusion, which draws from the results of 14 other similar recorded industry water hammer events. In addition, the VY engineering staff concluded a preliminary comparative stress analysis which demonstrated adequate margins to system pipe failure under the postulated water hammer transient loading. The inspector conducted a walkdown of the accessible RCIC and HPCI piping and noted that the HPCI vacuum breaker line supports in the vicinity of the containment penetration were rigid supports. This observation was communicated to the VY engineering staff for follow-up.
An inspector follow-up item (IFl 97-12-03) will track the licensee's final resolution of this issue and ensure inspector review of the corrective action. On January 30, the VY Plant Operations Review Committee and Plant Manager reviewed and approved a Basis for Maintaining Operation (BMO No. 98-01) for this issue, titled, "Effect of HPCl/RCIC Va,:uum Breaker Location on Water Hammer." The inspector reviewed BMO N > 98-01 and found it to satisfy the requirements specified in the licensee's BMO t.wid61ines (administrative control document). The operability determination was clearly and concisely written and the corrective action plan appears to be appropriate. The corrective action item to implement a multi-
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disciplined evaluation team intended to formulate a resolution to the vacuum breaker location issue.
c.
Conclusions VY's identification and initial corrective actions to address the potentially adverse location of the HPCI and RCIC system exhaust line vacuum breakers outside the torus air space were appropriate. The operability determination was reasonable and the associated BMO clear, concise, and appropriately founded. The licensee's 10 CFR 50.72 notification was timely and consistent with the established NUREG 1022 guidance. An inspector follow-up item (IFl 97-12-03)will track licensee resolution of this potentially adverse design issue.
E2 Engineering Support of Facilities and Equipment E.2.1 Review of Basis for Maintainina Ooerations (BMOs) Proposed by VY to Remain Open Followina the 1998 Refuel Outaae a.
Backaround and Inspection Scoce (71707,37551,40500) During this inspection period, the inspectors examined all of the active BMOs and the associated corrective actions which the VY staff proposes to remain open I following the completion / restart from the 1998 refuel outage. The purpose of this examination was to conduct a preliminary assessment of the licensee's corrective actions.
The list of active BMOs totaled 57, as of January 2,1998, with 6 pending final review and approval. In recent months, the VY staff has periodically (every two weeks) reviewed all of the BMOs and updated their plans for BMO closure. The licensee's January 2,1998 summary report reflected a significant re-assessment and resultant change to the total number of BMOs that are projected to remain open through the 1998 refuel outage. The inspectors' preliminary assessment follows: b.
Observations and Findinas Torus issues BMO 96-05, Torus water temperature limit
in November 1995, the licensee identified that the TS limit for torus water temperature (100 degrees F) could adversely affect emergency core cooling system (ECCS) pump net positive net positive suction head (NPSH) limits post-LOCA.
Accordingly, the VY staff administratively reduced the torus temperature limit to 90 degrees F pending further analysis and the submission of a TS amendment. The inspector noted that at the conclusion of this inspection period the TS amendment had not been submitted. The inspector determined that the licensee proposes to leave this BMO open, pending NRC staff review and approval of their pending TS chang,.g
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- BMO 97-15, Torus rupture disc / EOP design basis
in January of 1997, during the annual Emergency Preparedness exercise, the VY j staff identified that, under specific conditions, the Emergency Operating Procedures I (EOPs) direct containment flooding which could potentially have the consequence of actuating the torus vent system rupture disc. The potential rupture disc actuation results from the combination of hydrostatic pressure and containment pressure, as .; determined by analysis. The licensee has been operating with the rupture disc i isolated, via a closed downstream isolation valve, since this issue was discovered.
The inspector's review of BMO 07-15 and the January 2,1998 status report identified that VY does not plan to focus engineerir,g resources on this issue until after the 1998 refuel outage.
BWlO 97-30, Torus water temperature uncertainties
' The torus water temperature TS and analytical limit, used m the safety analysis, are the same value. However, the safety analysis does not account for the actual instrument uncertainty of 13*F. To compensate for this temperature instrument uncertainty, the VY staff has re-calibrated the torus temperature instruments to raad j 3*F higher than actual temperature to ensure the analytical safety margins are maintained. The inspector deterrained that final resolution of this instrumentation uncertainty issue awaits engineering resource commitment, currently scheduled for after the 1998 refuel outage.
BMC 97-41, Torus level instrument uncertainties -
The licensee identified that, in addition to the torus temperature instrument uncertainty, the torus level instrument uncertainty was not appropriately taken into account in maintaining torus water level within the allowable band. As a consequence, control room operators must maintain a narrow torus water level j band (within one inch of the upper and lower limit) to ensure adequate down comer submergence and torus water inventory for post-LOCA pressure suppression and heat removal. The inspector determined that modification of the level instruments during the 1998 refuel outage has not been scheduled.
Alternate Coolina System issues BMO 97-04, Alternate cooling system TS
The TS for alternate cooling system (ACS) treats the system as equivalent to one service water subsystem. However, the design bases of the ACS do not accommodate accident heat loads, and the system is not single failure-proof. The licensee has not submitted a TS amendment to correct this issue.
i l Appropriate administrative controls are in place. However, the inspector determined that closure of this BMO is scheduled for after the start-up of the 1998 refeni outag. .
BMO 97-43, Alternate cooling system (ACS) cooling tower fan bushing
The ACS cooling tower fan 2-1 motor torque, during fan startup, exceeds the torque rating of the fan bushing. The licensee does not consider the outage an opportunity to resolve this issue and the corrective maintenance is targeted to be performed on-line following unit restart.
BMO 97-52, Locs of alternate cooling system inventory through non-seismic pipe
break The licensee has determined that a rupture in a non-seismic portion of the service water system could result in a loss cf water from the alternate cooling system deep l basin. The FSAR Section 10.8 states that seismic effects will not impair the ability of the ACS to accommodate on-site water storage requirements in the event of a loss of the Vernon Pond, for one week. Additional preliminary analysis indicates that the ACS piping should not catastrophically fait during a seismic event.
BMO 97-56, Alternate cooling system winter operation
The present calculation for cooling tower thermal performance does not account for reduction of air flow due to icing. Also, icing may increase the fan motor power requirements.
Emeroency Diesel Generators BMO 97-39, EDG support system pipe welds, (Updated) Inspector Follow item (97-
08-01), Follow-up of industry operating experience identified partial penetration welds in skid mounted EDG support system piping (jacket water cooling and lubricating oil piping). The specific case of VY's EDG is currently being reviewed by the NRC.
This apparent non-conforming EDG support piping issue appears to be generic in nature. The inspector reviewed the licensee's current operability assessment tmd concluded that it is consistent with Generic Letter 91-18, revision 1, guidance. The licensee completed a detailed stress analysis and seismic evaluation of the EDG support piping which was provided to the NRC staff for review. Long-term resolution of these apparent non-conforming EDG support piping configurations is currently being evaluated by the licensee. The current EDG operability assessment is appropriate, but long-term resolution of this probiem is pending licensee action and additional NRC staff review. Inspector follow item IFl 97-08-01 remains open.
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BMO 97-46, EDG fuel oil trans.ar cross-connect
This BMO identified that the FSAR states that a single active failure in the EDG fuel oil transfer system will not result in the loss of either EDG, because operators will have three hours to cross-connect to the other EDGs day tank supply system.
However, there was no procedure to acesmplish this, and, given that the first indiation of a problem would be the day tank low level alarm, operators would actuah/ have less than one hour to take action. However, this issue has been compenced for, in the interim, by proceduralized operator actions to fill the affected dai tank using a portable pump.
BMO 97-51, Failcre position of EDG day tank level control AOVs
VY identified that a loss of instrument air causes the EDG fuel oil day tank fill line AOVs to fail closed. Although this is compensated for by proceduralized operator actions, it is contrary to the FSAR statement that each EDG will have a continuous supply of fuel and that fuel oil makeup is cccomplished automatically.
IPEEEissues BMO 97-14, Reactor building flooding from firs system pipe break
On April 8,1997, the licensee concluded that a break in the non-seismic fire system piping in the reactor building has the potential to flood one of the RHR corner rooms, if the water level got high enough in the reactor building, the water could potentially compromise electrical equiprnent required for safe shutdown. The inspector verified that this potential concern has been compensatec; for by VY via manual operator action in existing emergency and off-normal operating procedures.
However, long-term resolution has been targeted by VY to be consistent with their IPEEE completion schedule.
Although this issue was discovered as a result of the iPEEE project, the inspector J has determined that the resolution of this potentially degraded /non-conforming condition should be addressed in accordance with 10 CFR 50, Appendix B, per the guidance of Generic Letter (GL) 91-18, Revision 1. The inspector noted that the issue stems from a statement in the 1988 VY flooding report, which indicates that fire system flooding could be successfully hcrdled. Also, the flooding design basis requires that safety-related equipment required for safe shutdown be protected from flooding due to the postulated rupture of non-seismic piping.
BMO 97-23, Reactor building and Administration building flooding from non-seismic
systems The licensee determined on May 20,1997, based upon further review of BMO 97-14 and IPEEE efforts, that breaks in the non-seismic service water or fire system piping in the reactor building and administration buitriing have the potential to adversely affect electrical equipme.nt required for safe shutdown. The licensee compensated for these additional potential non-seismic piping breaks wit'i similar
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manual operator actions via existing emergency and off-normal operating procedures. As in BMO 95-14, the licensee's resolution of this issue is proposed to be consistent with their IPEEE schedule.
BMO 97-44, Tornado missile protection
' As a result of the Architect Engineering Inspecibn (97-201), certain safety related equipment was identi9ed to potentially have inadequate tornado missile protection (i.e., service water supply to the traveling screens, EDG fuel oil transfer lines, service water to the EDG, and the EDG exhausts). In addition to the basis for operability of the identified systems, BMO 97-44 targets the resolution of this issue j with the completion of their IPEEE activities.
The inspectors identified that BMO 97-44 stato that VY was not subject to the i GDC requirement * for protection from the effects of tornadoes, due to the age of the plant. However, VY describes its conformance to these GDC in their FSAR.
EQ lssues BMO 97-06, Switchgear room EQ concerns following a High Energy Line Break
(HELB) On March 10,1997,the licenses postulated that following a HELB in the turbine building, there is a potential for steam intrusion into the switchgear room via failure of the interfacing turbine building block wall and/or steam flow through the ventilation system. Thi.: postulated accident scenario was viewed as a potential non-conforming condition, in that, the switchgear room is defined as a mild environment within the EO Program. The inspectors verified the adequacy and implementation of the licensse's temporary modifications (interim corrective action) to preclude en minimize the impact of the postulated accident. The inspectors determined that final resolution of this issue is targeted for post-restart from the ? 998 refuel outage.
BMO 97-11, Effect of steam tunnel and turbine building HELBs on the HVAC rooms
The potentially adverse safety issue identified in BMO 97-11 resulted from the licensee's 'ollow-up actions to their March 7,1997 ENS notification (Event No.31915). Specifically, the licensee identified that there is a lack of documentation to support the EQ assumptions that the HVAC equipment rooms remain a mild en6onment for a main steam tunnel or turbine building HELB. The over-pressure conditions created by these HELBs could cause failure of the HVAC masonry walis, thus comcromising the EQ rating of equipment contained within.
The licensee implementeo compensatory actions (manual operator action) via existing procedures, pending proposed changes to the EQ progra. .
Miscellaneous issues BMO 97-28, Appendix J weaknesses for HPCI suction
This BMO identified that a single failure could cause both HPCI suction MOVs from the torus to open. This potentially results in the condensate storage tank (CST) suction line check valves (HPCI has a common suction line with upstream isolation valves from either the torus or CST) to act as the containment boundary. However, the current Appendix J program lists the containment boundary as the water seal in the torus and the torus suction MOV. The inspectors determined that the resciution of this issue requires a design change which the licensee states will not be completed prior to restart from the 1998 refuel outage due to time constraints.
BMO 97-50, Reg Guide 1.97 instrument safety classification
A number of RG 1.97 instruments were not maintained at the appropriate safety class. The inspectors, found that the licensee has provided a sufficient operability basis, however, VY does not consider the 1998 refuel outage an opportunity to resolve this issue.
BMO 97-53, Qualification of reactor building doors
The licensee identified that the reactor building doors were purchased non-nuclear safety (NNS), but have a safety class (SC) 2 function. The inspectors found that the licensee has provided a sufficient operability basis, however, VY does not i consider the 1998 refuel outage an opportunity to resolve this issue.
) BMO 97-58, Pitting of service water strainers
The licensee identified that the shells of the two service water strainers have i localized areas of wall thinning that are less than the ASME Code required minimum wall thickness. The inspectors found that the licensee has provided a sufficient operability basis, however, VY does not consider the 1998 refuel outage an opportunity to resolve this issue.
BMO 97-01, RBCCW water hammer
The licensee determined that certain LOCA scenarios could produce conditions that could result in a wau hammer event in the reactor building closed cooling water (RBCCW) system piping inside the drywell (reference GL 96-06). A docketed request for deferral by the NRC staff had not been made as of the conclusion of the inspection period.
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Conclusions Based upon the inspector's review of active Basis for Maintaining Operations (BMOs) and the VY staff's current projections for closure of BMOs following the Spring 1998 refuel outage, it appears that the schedule for completion / resolution of a number of th3 BMOs may not satisfy the intent of Generic Letter 91-18, Revision 1, and 10 CFR 50, Appendix B. As a result, a management meeting was scheduled to further review the licensee's actions. The inspectors will monitor the licensee's progress at resolving these non-conforming conditions. (IFl 97-12-04) E8 Miscellaneous Engineering issues E8.1 IClo_ged) Unresolved item 97-05-03: Seismic desian considerations for the
volt DC oower system Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," delineates variables to be monitored duri.3g the course of an accident. The inspector reviewed Table 1, "BWR Variables," of Regulatory Guide 1.97 and determined that none of the variables monitored by the process liquid radiation monitoring system were included in the table. The inspector concluded that there was no regulatory basis that required the process liquid radiation monitoring system at VY to be classified as a safety-related system. Therefore, the i24 VDC power system is not required to be safety class electrical, and the batteries need not be seismically mounted.
The licensee performed a review of the FSAR to verify that the information on the i24 VDC power system was accurate. They noted that the original design specification for the batteries stated, "the racks shall be equipped with retainers and spacers (generally known as earthquake protection equipment)." VY indicated that the retainers are thought to refer to the steel that surrounds the batteries, and that the spacers to refer to a foam material that would be installed between the jars to provide lateral support during a seismic event. The original batteries were replaced in 1984. From review of that activity, VY determined that the existing racks are original equipment. Spacers were not installed during the replacement (which is the current configuration), and it is not known whether they were in place with the original batteries. VY generated an internal commitment to review the wording of the statement in the FSAR that refers to the seismic design of the racks.
As stated in the definitions section of the FSAR, only safety systems will have safety design bases included in the FSAR system descriptions. Given that the process liquid radiation monitoring system is not required to be a safety system, inclusion of a safety design basis description does not appear to be appropriate. VY continues to review this section of the liquid process radiation monitoring system description in the FSAR for revision or deletion.
The inspector concluded that thei24 VDC power system is not a safety class electrical system, therefore the batteries need not be seismically mounted. There
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was no violation of requirements. Imprecise wording and inconsistencies in formatting in the FSAR which led to this issuc being raised are being addressed by the licensee. Accordingly, unresolved item URI 97-05-03 is closed.
E8.2 (Closed) Unresolved item 97-05-04: Interface of a retired-in-clace system with the j hiah oressure coolant iniection system ' a.
Backaround and Inspection Scope (92903) The torus /drywell pumpback system was an original pisnt system that was used to establish differential pressure between the torus and the drywell. When the containment nitrogen inerting system was subsequently installed, the torus /drywell pumpback system was no longer required, and was retired in place The system is still connected to the high pressure coolent injection (HPCI) gland exhauster discharge line to the standby gas treatment iSBGT) system; however,it is
apparently not seismically qualified, and isdation from the HPCl/SBGT line is provided by the retired pumpbacle unit. In addit'on, the piping and instrumentation diagrams (P&lDs) for these systems provide conf'leting information on the safety classification of the pumpback system piping, the NPCl/SBGT line, and location of safety class breaks.
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Observations and Findjngi The inspector reviewed event report (ER) 97-0821 which VY had generated to address issues associated with the pumpWek system. The VY investigation determined that the HPCl/SBGTline (3"UPCI-13) had been recias*<ified as a safety class 3 line in 1988, but that the attached lines had not been addressed. The ER indicated that the remaining aedons to resolve this issue were to indh ete the safety class designation of the interfacing lines on the applicable drawings, and to verify the seism!c integrity of the interfacins lines.
The inspector was concerned that a potentic:ly degraded /non-conforming condition was not being appropriately addressed by VY. Specifically, as submitted in Juris 1997, the Eft Jn6cated tict, based on incoraistencier, in the drawings, it is possible that th6 T'HPCl-13 could he safety class 3 (SC3) without being sefsmically analyzed. It is possible ihat an apparernly non-seismic line to the pumpback blower is un'sola' ale from the pctentially SC3 HPCI pi nd see) ex.hauster disaharge line."
t The irsspector found the initial opability determirwtion M ta edequate. However, the inspector arked why cordnWed operation whilo Mc issu1s were being resolved had not besn aEressed through the bas:s fm rnainttining operation (BMO) process. The lichasee determined that 3"MCCI-13 was SC$, bb+ that the safety classification of the interfacing lines tuas indeterminate, wWh egaie, appsured to warrant a BMO. The inspector discussed this concern with \\'Y mar.apijment. VY determined ticat a BMO was not required, g!ven that se'6mh adeqtacy had beef, established by walkdowns as part of the opsrability determiration, em that no compensatory measures were required as a result of the conditbn.
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im . Seismic and safety grade classification issues associated with the HPCI gland exhauster discharge line to the SBGT system, the torus /drywell purnaback system, and interfacing connections, wera being appropriately addressed. The seismic classification of these as-built interfacing piping systems were determined to have been appropriate since original construction. No violations of NRC requirements were identified. Therefore, this unresolved item (URI 97-05-04)is closed.
E8.3 (Closed) Unresolved item 97-03-02: Electrical cable seoaration issue a.
Backaround_and Insoection Scoce (92700,92903) .
As previously documented in inspection reports 50-271/97-03(section E1.1) and 97-05 (section E1.4), the VY staff identified and implemented corrective actions to address their discovery of a number of original construction electrical cable and j instrumentation control wiring separation discrepancies. The adequacy of the ' licensee's initial response to this issue was documented in inspection reports 97-03 and 97-05 and found to have been acceptable. The unresolved item was left open pending inspector review of the adequacy of longer term corrective actions and pending assessment of the issue with respect to appropriate enforcement action, because of the identified non-compliances with the VY Updated Final Safety i Analysis Report (UFSAR). The inspector examined the licensee's corrective action plan for the electrical cable separation issue and monitored the licensee's tracking to. completion of two remaining action items.
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Observations and Findinas The inspector determined that VY had updated the Basis for Maintaining Operation (BMO) 97-13, " Separation of NNS Power and Control Cables," with revision 3, dated October 23,1997. There are about 60 cables affected by this BMO.
Revision 3 retracted ' Safety Evaluation No. 9714 (reference inspection rep 9rt 50-i 271/97-05)and updated the BMO action pian to reflect the resolution of two NNS power cables end eight NNS control celotes by reronting the 10 afiected cables Revision 3 also assigned the Design Etigineering staff responsibility to implement and complete the necessary dodgn chenges to ercure the remaining NNS cabjes identified in the BMO comply with the VY liceasing basis (VY Specidcation 027 and UFSAR, Section (t4.6) prior to tre complation of the 1998 refuelit;g outage.
The inspector er.amined BMO 9713, revision 7, the current Major Project Work List (MPWL), and discussed the status af the pending design changes with the responsible rductrical design ergineering manager. The manager conCrmed that resources were currently employed to resolve the cable separation issue pr' ice to ctart-up from the 1998 refuel outage. The manager also confirmed that a comprehentive reverification effort to examine cehle separatim in vertical run cch raceways routed through non-designated manways was underway and that these activitkes were being tracked on the MPWL.
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As a result of the cable separation issue corrective actions, the licensee identified additional similar cable separation design discrepancies, as documented in LER 97-21, dated November 5,1997. To address the potentialimmediate system operability impact, the VY staff initiated BMO 97-16 and developed a corrective action plan. On October 23,1997, Revision 1 to BMO 97-16 was issued to discuss the cable separation concerns associated with the alternate cooling system - (ACS) tower fan CT-2-1 cnd to amend the BMO with'a new safety evaluation (SE No. 97-28, revision 1) to support tFe compensatory actions for the CT-2-1 fan supply breaker configuration changes. The inspector examined these corrective actions to verify their implementation and adequacy. Other remaining work activities including: the formal closure of BMO 97-16; the CT-2-1 fan cable separation problem and the revision of the electrical separation criteria; and, corrective actions to reflect the ACS cabling anomaly were verified by the inspector to be appropriately scheduled for completion in the licensee's corrective actions program. As previously discussed, all safety and non-safety designated manways and their associated vertical run cable raceways will be examined and discrepancies addressed, prior to unit restart from the 1998 refuel outage.
The VY staff's identification and corrective actions to address the above stated electrical cable separation "old design issues" were appropriate and were executed, i I or planned to be accomplished, in a time period commensurate with their safety ( significance. These actions are consistent with Generic Letter 91-18 and 10 CFR {~ l 50, Appendix B, Criterion XVI, " Corrective Actions." The inspector also observed that the reporting of these issues was consistent with 10 CFR 50.72 and 50.73 requirements as documented in LER No. 97-006, Supplements 1 and 2, and LER No.
97-021. The two referenced LERs and supplements were reviewed on-site by the resident inspectors, c.
Conclusions The failutc to implenient appropriate electrical cabling separation is a violation of 10 CFR 50, Appendix 8, Criterion Ill, Design Control, as well as the VY UFSAR, but appears to have been the result of inadequate original design specifications and . installation practices. This violation was identified by the VY staff as a result of a voluntary initiative, and the corrective actions were prompt and comprehensive.
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 97-12-05, LER 97-006 and Supplements 1 and 2, LER 97-021, and URI 97-03-02 are closed.
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l IV. Plant Support F2 Status of Fire Protection Facilities and Equipment F2.1 Automatic Self-Closino Fire Door Found Inooerable t a.
Insoection Scone (71750) !. During a routine plant inspection on January 13, the inspector noted that the closing mechanism for ?ne high pressure coolant injection (HPCI) room automatic l self-closing fire door was not properly set up.
L l . b. Observations and Findinas The inspector observed that the drop weight cable which provides motive force to l pull the HPCI room automatic self-closMg fire door closed was at of one of the ! pulleys and was resting on the pulley axN. The inspector reported the condition to ( the on-shift operations crew and the shift engineer (responsible for fire protection issues) confirmed that this was a discrepant condition. A closure test was j.
performed with the rigging in the as-found condition and the dou did not close.
After the cable had been placed back onto the pulley, the door was tested satisfactorily.
The HPCI fire door automatic closure function is tested semi-annually in accordance with surveillance procedure OP-4019," Surveillance of Plant Fire Barriers and Fire L Rated Assemblies." The door had last been tested in October 1997 and had operated properly. As a result of troubleshooting on January 13, VY considered the most likely cause of the cable coming off of the pulley was due to the door being j leaned on, which caused it to move back and forth. Due to the short run of cable i between the affected pulley and the counterweight, this caused one of the two ! clamps that secures the counterweight to the cable to enter the pulley and walk the l cable out of the channel. VY fire protection personnel observed this mode of failure on one occasion during troubleshooting. The clamp was subsequently repositioned to below the second clamp. In addition, the fire protection eng neer is verifying that . the cable is properly engaged in the pulley during performance of the monthly fire protection surveillance.
l The inspector reviewed the Vermont Yankee Safe Shutdown Capability Analysis, revision 5, dated November 19,1996. Section 2.5, " Safe Shutdown Systems," ider tifies the HPCI system as a safe shutdown system. Section 3.0, " Determination of Fire Areas and Fire Zones," indicates that the wall between the HPCI room and the torus room is a fire barrier in accordance with Appendix A of Branch Technical Position APCSB 9.5-1, " Guidelines for Fire Protection for Nuclear ' Power Plants." The inspector concluded that the inoperable HPCI fire door constituted a violation of Technical Specification 6.5.A, " Plant Operating Procedures," in that the HPCI fire door was not maintained operable in accordance with VY's' safe shutdown capability analysis. (VIO 97-12-06)
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- Conclusions The HPCI room automatic self-closing fire dooi was inoperable for an indeterminate period between October 2,1997 and January 13,1998. VY took prompt action to restore operability when identified by the inspector and to correct the suspected j cause. The inoperable fire barrier was a violation of 10 CFR 50 Appendix R requirements. (VIO 97-12-06) V. Management Meetings X1 Exit Meeting Summary The resident inspectors met with licensee representatives periodically throughout the inspection and follouing the conclusion of the inspection on February 13,1998.
At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the preliminary inspection findings.
X3 Review of Updated Final Safety Analysis Report (UFSAR) A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the area inspected. The inspectors verified that the UFSAR wording was consistent with the observed practices and procedures and/or parameters.
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INSPECT!ON PROCEDURES USED 37551 Onsite Engineering 40500 Effectiveness of Licensee Controls in identifying, Resolving, and Preventing Problems 61726 Surveillance Observations 62707 Maintenance Observations 71707 Plant Operations 71714 Cold Weather Preparations 71750 Plant Support Activities 92700 Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities 92902 Followup - Maintenance j 92903 Followup - Engineering
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22 ITEMS OPENED, CLOSED, AND DISCUSSED OPEN ! IFl 97-12-01 Instrumentation loop analytical and TS allowable limit concern.
IFl 97-12-02 Post-LOCA torus temperature determined to be higher than previously analyzed IFl 97-12-03 Potential condition outside design basis involving location of vacuum breakers for the HPCI and RCIC steam exhaust lines ! IFl 97-12-04 BMOs proposed by VY to remain open following the 1998 refueling outage ! NCV 97-12-05 Failure to implement appropriate electrical cabling separation i VIO 97-12-06 Inoperable HPCI room fire barrier CLOSED l URI 97-05-03 Seism;c design considerations for the 24 volt DC power system i URI 97-05-04 Interface of a retired-in-place system with the high pressure coolant injection system l l NCV 97-12-05 Failure to implement appropriate electrical cabling separation URI 97-03-02 Electrical cable separation issue l URI 97-02-09 Instrumentation loop analytical limits exceed the TS allowahle limit l LER 97-006, Use of an inadequate design. nplementation document during initial 97-006 Supp1, 2 construction results in the failure to maintain proper electrical . separation of electrical cables r l ! LER 97-021 Division SI powered and Sil powered cables located in same manhole contrary to FSAR statement and Vermont Yankee separation criteria l due to inadequate original design specifications DISCUSSED IFl 97-08-01 Safety grade qualification of welds in emergency diesel generator support systems !
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PARTIAL LIST OF PERSONS CONTACTED G. Maret, Plant Manager F. Helin, Tech. Servi es Superintendent . E. Lindamood, Director of Engineering K. Bronson, Operations Manager M. Watson, Maintenance Superintendent M. Desilets, Radiation Protection Manager R. Gerdus, Chemistry Manager G. Morgan, Security Manager l l a i I
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LI3T OF ACRONYMS USED BMO Basis for Maintaining Operation CFR Code of Federal Regulation CR control room CS core spray EDG emergency diesel generator ER Event Repon GL Generic Letter HPCI high pressure coolant injection IFl ' Inspector follow item IN Information Notice KV kilovolt ' LCO Limiting Condition for Operation LER Licensee Event Report LPCI low pressure coolant injection MCC motor control center NRC Nuclear Regulatory Commission NNS Non-nuclear safety , ' PORC Plant Operations Review Committee QA Quality Assurance-RHR residual heat removal RP radiation protection TS Technical Specifications j UFSAR Updcted Final Safety Analysis Report j URI unresolved item VAC volts alternating current VDC volts direct current VY Vermont Yankee l ' , }}