IR 05000271/1999006
| ML20216F336 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/13/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20216F321 | List: |
| References | |
| 50-271-99-06, NUDOCS 9909210263 | |
| Download: ML20216F336 (16) | |
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, U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket N l Licensee N DPR-28 Report N Licensee: Vermont Yankee Nuclear Power Corporation Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: June 21 - August 1,1999 Inspectors: Brian J. McDermott, Senior Resident inspector Edward C. Knutson, Resident inspector Russell J. Arrighi, Resident inspector, Pilgrim NPS Approved by: Clifford J. Anderson, Chief
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Projects Branch 5 Division of Reactor Projects
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9909210263 990913 PDR 0 ADOCK 05000271 PDR ,
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EXECUTIVE SUMMARY Vermont Yankee Nuclear Power Station NRC Inspection Report 50-271/99-06 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six week period of routine activities by the resident inspectors assigned to the facility. In addition, a review of the reactor core isolation cooling system was performed by a resident inspector from another Region i facilit Operations
. Appropriate control of safety system alignments, implementation of Technical Specification required actions, and adequate operability reviews for degraded equipment were observed during routine control room tours. (Section O1.1)
. The NRC identified two service water leaks from room cooling unit coils. Although the leaks did not render any equipment inoperable, they were not detected by W's monitoring during a chemical treatment designed to remove mircobiologically induced corrosion (MIC) from the service water piping. (Section 01.2)
. The reactor core isolation cooling system was properly aligned to support system operability and no concerns were identified during an NRC walkdown of the syste (Section O2.1)
Maintenance !
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. The observed maintenance activities were performed well. Good radiological protection !
department support was noted during work on two main turbine stop valves located in a i high radiation area. (Section M1.1)
. Surveillance activities observed during this inspection were performed well. W's early I identification of degradstion on several AS-2 battery cells demonstrated a good attention I to detail during the routine surveillance. All of the surveillance criteria were met and the AS-2 battery remains operable. (Section M1.2)
. The NRC identified several material deficiencies on the two emergency diesel generators. Although the deficiencies had not affected operability, they collectively demonstrated that prior maintenance activities were not well controlled and that routine ;
monitoring of the equipment's condition was not sufficient. Individual corrective actions have been completed and W is evaluating a systemic approach to prevent repeat problems. (Section M2.1)
Enaineerina
. W identified that an existing design basis calculation for the torus-to-reactor building vacuum breakers was inconsistent with the current plant configuration. Initial W il
Executive Summary (cont'd)
evaluation concluded that this inconsistency could have created a condition outside the plant's design basis, and the issue was conservatively reported to the NRC under 10 CFR 50.72. Pending re-analysis, W demonstrated that the vacuum breakers were operable under the existing plant conditions. W subsequently concluded that the vacuum breakers met their design basis under all conditions, and the initial NRC notification was retracted. (Section E1.1)
However, W's failure to revise the limiting case analysis for containment depressurization to reflect a design change (during plant construction) was a violation of 10 CFR 50, Appendix B, Criterion Ill, " Design Control." This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. The issue was entered in W's corrective action program as ER 99-0773.
- W engineering effectively supported plant operations by identifying measures to reduce the operational impact from high ambient temperatures during the summer of 199 (Section E2.1)
Plant Suooort
. Several conditions that could have spread contamination or indicated a change in radiological conditions were identified during routine NRC plant walkdowns. Although no actual contamination issues resulted, the inspector considered that the conditions had likely existed long enough to have previously been identified by W plant personne (Section R1.1)
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TABLE OF CONTENTS EXECUTIVE SUM MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii l TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ................... 1 1. Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 1 01' Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Observation of Routine Plant Operations . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Service Water System Monitoring During Chemical Treatment . . . . . 1 02 Operational Status of Facilities and Equipment .. .................... 3 02.1 Safety System Walkdown - Reactor Core isolation Cooling System . . 3 11. M aintena nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 o M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . .. .............. 4 l M1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l M1.2 Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . . . . 5 l'
M2.1 Material Condition of the Emergency Diesel Generators . . . . . . ... 5 I ll . Enginee ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 6 E1- Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 '
E Design Basis Calculation for Torus Vacuum Breakers . . . . . . . . . . . . 6 l E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . 8
! E2.1 Review of Actions to Address High Ambient Temperature . . . . . . . . . . 8 l
IV. Plant Support . . . . . . . . ................................................... 9 l
l R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . 9 R Contamination Control Practices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 i V. M anagement Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l X Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 ATTACHMENTS Attachment 1 - List of Acronyms Used Attachment 2 - ltems Opened, Closed, or Discussed iv
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Report Details (
t Summary of Plant Status Throughout most of the inspection period, the Vermont Yankee (VY) plant was operated at 100% power. Power reductions were made on June 23, July 15, and July 26, for control rod pattem adjustments. On July 13, power was reduced and single loop operation was necessary to support planned maintenance on the "B" recirculation pump motor generator. Minor power reductions were also made to support routine surveillance testing during this perio l. Operations 01 Conduct of Operations'
01.1 Observation of Routine Plant Operations Inspection Scope (71707)
The inspectors routinely toured the control room to assess the conduct of activities, verify safety system alignments, and verify compliance with Technical Specification (TS)
requirements. Equipment deficiencies identified in control room logs were reviewed and discussed with shift supervision, to evaluate both the equipment's condition and the adequacy of VY's initial response to the issu Observations and Findinas No problems were identified with the status of plant safety systems during the control ;
room tours or review of Event Reports (ERs). A sample review of work orders and ERs a found that the basis for operability of degraded equipment was adequately evaluated and documente Conclusions l
Appropriate control of safety system alignments, implementation of Technical Specification required actions, and adequate operability reviews for degraded equipment j were observed during routine control room tour .2 Service Water System Monitorina Durina Chemical Treatment !
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' Insoection Scope (71707)
From June 18 to July 30, VY performed a chemical cleaning of the service water (SW) i and residual heat removal service water (RHRSW) systems. This activity was part of an I integrated plan to deal with performance degradation caused by microbiologically l Induced corrosion (MIC) in these systems, as discussed in inspection report i
' Topical headings such as Oi, 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
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50-271/99-01. The inspector examined W's actions to monitor system performance during the chemical treatmen b. Observations and Findinas To remove MIC from the piping systems normally supplied with river water, a penetrating bio-dispersant was added to the suction bays for the service water pumps. The bio-dispersant is designed to slowly remove small particles of MIC, suspend the solids, and allow them to be flushed through the system without any accumulation. No detrimental affect on system operation was anticipated by the licensee. Monitoring of this activity consisted of measuring the amount of suspended solids in the SW discharge every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> !
The treatment was planned for 30 days and was a one time, high concentration treatment. The inspector noted that the licensee provided additional management oversight for this first-of-a-kind evolution. During the pre-job briefing, the only recommended in-plant monitoring for this activity was a heightened sensitivity to service water system leaks. The licensee stated that other facilities had experienced some minor service water leaks after the bio-dispersant treatmen The inspector considered that monitoring for leakage and evidence of heat exchanger blockage of SW cooled components would nonetheless be appropriate during the chemical treatment. The inspector discussed this concern with W management. W's position was that the chemical treatment had been performed successfully at several other nuclear power plants, and that they considered monitoring for suspended solids to be adequat On July 7, the inspector identified a SW leak from a safety-related room cooler in the "A" ECCS comer room (contains the "A" loops of the residual heat removal (RHR), RHRSW, and core spray (CS) systems). The affected room cooler, reactor recirculation unit (RRU) 7, is required to be operable for the ECCS pumps in the room to be considered operable. W engineering determined that RRU-7 could be considered operable as long as the SW leak rate was less than 3 gpm. The measured leak rate was less than gpm, so operability was not an lmmediate issue. W is continuing to monitor the leak rate, and plans to replace the heat exchanger during the upcoming refueling outag On July 16, the inspector identified a SW leak from a room cooler in the feedwater pump room. The affected cooler, TRU-4, is not a safety-related component. A work request was generated to repair the leak. After the close of this inspection, W identified a SW leak from another feedwater pump room cooler, TRU- The causes of these SW leaks had not been identified at the close of the inspection period. While they may be unrelated to the SW chemical treatment, the inspector considered that they were the types of failures that would most likely occur as a result of corrosion removal (that is, small leaks from thin-wall piping). In any case, directed monitoring of the SW system during the chemical treatment would have increased W's opportunities to have self-identified these problem .
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c Conclusions The NRC identified two service water leaks from room cooling unit coils. Although the leaks did not render any equipment inoperable, they were not detected by VY's monitoring during a chemical treatment designed to remove mircobiologically induced corrosion (MIC) from the service water pipin O2 Operational Status of Facilities and Equipment O2.1 . Safety System Walkdown - Reactor Core isolation Coolina System Insoection Scope (71707)
The inspector walked down accessible portions of the reactor core isolation cooling (RCIC) system to verify proper system configuration. The applicable surveillance procedures were reviewed to ensure they accurately captured the Technical Specification (TS) surveillance requirements. In addition, the inspector reviewed open operability evaluations, event reports, and maintenance requests to verify that any degraded conditions had been assessed for potential impacts on system operability, Observation and Findinos The inspector determined that the RCIC system was operable based on a system walkdown and review of applicable procedures, drawings, outstanding deficiencies, and the system's description contained in the Updated Final Safety Analysis Repor The inspector reviewed procedure OP 4121, " Reactor Core Isolation Cooling System Surveillance," against technical specificatiors (TS) 4.5.G and verified that the surveillance requirements were properly captured in the procedur The inspector noted that the RCIC flow test is performed above the pressure specified in TS (normal reactor operating pressure), but not to the pressure specified in the licensee's design basis document (DBD). The DBD specifies that the system is designed to provide flow at a pressure equivalent to the lift set point of the lowest safety relief valve. A review of completed surveillances revealed that the system was tested at this higher pressure. Pending review of VY's addressing of the disparity between these two values for discharge pressure, this item will be tracked as an inspector followup item (99-06-01 Surveillance Test Value for RCIC Pump Discharge Pressure). Conclusions The reactor core isolation cooling system was properly aligned to support system operability and no concerns were identified during an NRC walkdown of the syste .
. 4 11. Maintenance M1 - Conduct of Maintenance M1.1 Maintenance Observations Inspection Scooe (62707)
The inspector observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and Technical Specification requirements were satisfied, adequate measures were in place to ensure personnel
- safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testin Observations and Findinas The inspector observed all or portions of the following maintenance activities:
. Main turbine stop valves 3 and 4 test solenoid repairs, observed June 23 During quarterly main turbine stop valve testing on June 22, the test solenoid faile Troubleshooting and repair was performed in a high radiation area. Prejob briefings were clear and good supervisory support was noted. The inspector observed ge,od use of video and remote dose rate / exposure monitoring. Good planning and use of a low dose waiting area was effective in maintaining the workers' radiation exposure as low as reasonably achievabl .
"B" Recirculation pump motor generator brush replacement, observed July 13 No problems were noted during the brush replacement or the subsequent restart of the
"B" recirculation pump. The inspector observed that maintenance supervision was in the plant to support this activit , Conclusions The observed maintenance activities were performed well. Good radiological protection department support was noted during work on two main turbine stop valves located in a ,
high radiation are :
M1.2 Surveillance Observations insoection Scooe (61726)
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The inspector observed portions of a surveillance test to verify proper calibration of test l instrumentation, use of approved procedures, performance of work by qualified l personnel, conformance to Limiting Conditions for Operations (LCOs), and correct post- l test system restoratio !
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' Observations and Findinos a "A" and "B" EDG monthly surveillance, observed June 22 and 23 During the "A" EDG surveillance, the inspector observed that one of eight fasteners on a jacket water cooling flange was loose. The flange was not leaking, and maintenance personnel were able to tighten the fastener while the engine was running. Other EDG material problems are discussed in section M2.1 of this repor . CRD exercising, observed June 29 The inspector observed that the pre-test briefing was appropriately detailed for this routine surveillance. Good three-part communications were observed throughout the surveillanc . AS-2 Battery quarterly surveillance, observed July 27 W personnel identified pitting on the negative terminals on several cells during this surveiliance. Although this age-related condition does not immediately affect battery operability, it indicates that the cells will eventually need to be replaced. All of the surveillance test acceptance criteria were met and the terminal pitting observation was entered into W's corrective action program (ER 99-0883). Conclusions J Surveillance activities observed during this inspection were performed well. W's early
' identification of degradation on several AS-2 battery cells demonstrated a good attention to detail during the routine surveillance. All of the surveillance criteria were met and the AS-2 battery remains operabl M2 Maintenance and Material Condition of Facilities and Equipment M2s1 Material Condition of the Emeroency Diesel Generators insoection Scope (62707)
The inspector identified several material deficiencies associated with the emergency diesel generators.(EDGs). The inspector reviewed W's immediate and long term corrective action Observations and Findinos During routine walkdowns of the EDGs during this inspection period, the inspector identified several material deficiencies. For example, one of eight fasteners was loose on a jacket water cooling flange on the "A" EDG (discussed in section M1.2 of this report). In another instance. the inspector found that the bolting for the turbochargers on both EDGs were shorter than specified by the vendo .
In each instance, W acted promptly to address EDG operability and to enter the problems in the corrective action program. As a long-term response to these findings, W formed a multi-discipline team (i.e., mechanical, electrical, and structural) to perform detailed inspections of the EDGs. The plant manager stated that this team will also be used to perform walkdowns of additional safety significant plant systems. At the close of the inspection, the W team was in the process of developing their inspection pla The inspector determined that this was a good initiative and that it would likely reduce the number of equipment deficiencies in the immediate future. The inspector noted that a similar initiative was completed in 1997 (reference IR 50-271/97-05). However, the current initiative differs from the previous effort in that the team is collectively focused on a single system at a time, as opposed to the various disciplines independently inspecting large areas of the plant. While the new approach may prove more effective, the need for a repeat effort indicated that W's routine inspection activities have not been fully effectiv Conclusions The NRC identified several material deficiencies on the two emergency diesel i generators. Although the deficiencies had not affected operability, they collectively demonstrated that prior maintenance activities were not well controlled and that routine monitoring of the equipment's condition was not sufficient. Individual corrective actions have been completed and W is evaluating a systemic approach to prevent repeat problem !
111. Engineering E1 Conduct of Engineering E1.1 Desian Basis Calculation for Torus Vacuum Breakers Inspection Scope (37551)
On July 2, W reported a condition that was potentially outside the design basis of the plant (reference EN# 35890) to the NRC. During verification of the containment system design basis document (DBD), W identified that the analysis demonstrating that the torus to reactor building vacuum breakers would limit differential pressure to less than the design limit was not valid for the current system configuration. The inspector reviewed W's resolution of this issu Observations and Findinas The design function of drywell spray is to reduce drywell pressure under accident conditions by quenching steam. The use of dryweil spray is governed by W's emergency operating procedure !
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The design function of the torus-to-reactor building vacuum breakers is to allow air from the reactor building to enter the torus in the event that a negative pressure develops in I the torus. The system has two trains, each consisting of a passive check valve in series j l
with an air operated butterfly valve. Drywell-to-torus vacuum breakers (10 passive check valves) similarly prevent negative pressure from developing in the drywell. The differential pressure limit for both the torus and the drywell is 2.0 psid, with the pressure l inside the torus or drywell being less than pressure in the reactor building. Negative ,
containment pressure could occur as a result of a variety of normal operating and i accident condition The limiting scenario for negative containment pressure is an inadvertent initiation of containment spray during normal plant operation. In this scenario, pressure reduction is caused by cooling of the drywell atmosphere (as opposed to quenching steam in an accident scenario). Therefore, the greatest pressure reduction would occur when the greatest temperature difference existed between the spray water (which comes from the i torus) and the drywell atmosphere. For the " worst case" analysis, the design low l temperature limit of 50 F for torus water, and design high temperature of 165*F for drywell atmosphere, are used. If operation of the torus-to-reactor building vacuum breakers were not credited, the final torus pressure would be more negative than the allowable -2.0 psid. On the other hand, if the torus-to-reactor building vacuum breakers were assumed to open instantaneously, the final torus pressure would remain within the l design limi W's original calculation assumed that the vacuum breakers would essentially open instantaneously. That was a reasonable assumption for the original design l configuration, which consisted of only the single passive check valve. An air operated butterfly valve was added in series with the check valve to increase the reliability of the containment isolation function. This addition introduced a time dependency to the vacuum breaker function, since the air operated valve takes 10-15 seconds to ope However, the analyses were not updated to take the new time dependency into accoun It should be noted that containment pressure reduction is also a time dependent event; j
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that is, pressure will decrease over some period of time to a final, steady state value, which occurs when the drywell atmosphere temperature reaches equilibrium. The original analytical technique appropriately did not take time dependency into account, since the vacuum breakers were assumed to open instantaneously. While noting that i there would be some additional margin when taking this time dependency into account, l W made conservative decision to report this condition per 10 CFR 50.72, as a condition that was potentially outside the design basis of the plan In the interim, W was able to demonstrate, by the original analytical method, that the torus-to-reactor building vacuum breakers were operable (that is, drywell pressure would not be more negative than -2.0 psid under all conditions) as long as torus water temperature remained above 70*F and service water temperature remained above 33*F. Service water temperature is a factor because an inadvertent drywell spray event during power operations would require the RHR system to be in operation as an initial condition, and that would most likely be while operating in the torus cooling mod .
- 8 On July 30, W retracted the NRC event notification. The basis for retraction was that W indicated that they had completed a time dependent analysis which demonstrated that the negative containment pressure would be less than 2.0 psid for an inadvertent drywell spray event initiated under the worst case conditions of torus water and drywell atmosphere temperatur CFR 50, Appendix B, Criterion lil, " Design Control," states, in part, that, " . . . Design changes . . . shall be subject to design control measures commensurate with those applied to the original design . . ." Contrary to the above, on July 2,1999, W identified that a change to the torus-to-reactor building vacuum breaker system that was implemented during construction of the plant had not been subject to design controls commensurate with those applied to the original design, in that the limiting case analysis for containment depressurization had not been revised to account for the time dependency that the change introduced into the system response. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. The issue was entered in W's corrective action program as ER 99-0773. (NCV 99 05-02: Inadequate Design Control for Torus-to-Reactor Building Vacuum Breaker Modification) Conclusions W identified that an existing design basis calculation for the torus-to-reactor building vacuum breakers was inconsistent with the current plant configuration. Initial W evaluation concluded that this inconsistency could have created a condition outside the plant's design basis, and the issue was conservatively reported to the NRC under 10 CFR 50.72. Following re-analysis, W demonstrated that the vacuum breakers were operable under the existing plant conditions. W subsequently concluded that the vacuum breakers met their design basis under all postulated conditions and withdrew 'he NRC notificatio However, W's failure to revise the limiting case analysis for containment depressurization to reflect a design change (during plant construction) was a violation of 10 CFR 50, Appendix B, Criterion lil, " Design Control." This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC !
Enforcement Policy. The issue was entered in W's corrective action program as ER 99- )
077 i E2 Engineering Support of Facilities and Equipment E2.1 Review of Actions to Address Hioh Ambient Temperature Insoection Scoos (37551)
Weather conditions this summer have presented a challenge to plant operations. High temperatures and low rainfall have resulted in high river temperature. Consequently, the service water supply temperature has been close to its design limit of 85'F on several occasions. Because service water supplies cooling to a number of plant systems, other
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. 9 limits such as maximum torus water temperature were also approached. The inspector reviewed W's approach to coping with the high service water temperature, Observations and Findinos W developed several actions to reduce temperature instrument inaccuracy and thereby increase their margin to temperature limits. By measuring service water temperature locally at the service water supply header, the indication inaccuracy was reduce from 3 2.0*F to 1.0*F. By using an average of all available torus water temperature indications, )
W engineering demonstrated that the maximum indicated temperature for this parameter could similarly be increased by 1.0* By comparison of the temperatures of the torus room and torus water, W determined that the warm air in the room was contributing to torus heatup. A temporary modification, j TM 99-016, was developed and implemented to use a mobile air conditioning unit to cool the torus roo Other measures that were implemented for service water temperature reduction were to remove the floating boom at the intake structure to reduce heating of slow moving water, operating the main circulating water system in closed cycle when the service water supply temperature approached the temperature limit, and coordinating torus cooling operations with periods of high river flo These measures were effective in allowing the plant to continue to operate through the periods of high ambient temperature to date. In preparation for even higher temperatures, W engineering was preparing a basis for maintaining operation (BMO)
that would justify operation with service water temperature as high as 88'F. If the additional margin is not required this summer, VY plans to implement this increase as a change to the FSAR. W is also evaluating the possibility of a 3.0*F increase in the maximum allowable torus water temperatur Conclusions W engineering effectively supported plant operations by identifying measures to reduce the operational impact from high ambient temperatures during the summer of 199 IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Contamination Control Practices Inspection Scope (71750)
During this inspection period, the inspector noted several conditions that could have resulted in the spread of contamination, or which indicated potential inattentiveness to radiological condition l l
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b. Observations and Findinas in one case, packing leakage from a residual heat removal (RHR) system valve was collecting on the floor in a passageway; the condition had apparently existed for most of a shift, but no action had been taken to contain the leakage, survey the area, or restrict access, until pointed out by the inspector. The inspector subsequently noted that the equivalent valve in the other loop of RHR had absorbent material under it, apparently to contain packing leakage, but that the area was not posted as being potentially contaminated. In a third case, the inspector noted standing water around the cover of a potentially contaminated sump. The condition had existed for several days, but no action was taken to clean it up until it was pointed out by the inspecto Finally, the inspector noted that a contamination monitoring instrument that is permanently located outside one of the reactor building air locks was indicating on the order of 10 times its normal count rate. The inspector considered that the audible count rate should have clearly indicated an abnormal condition to any personnel trained in radiological controls; however, radiological protection (RP) department personnel were unaware of the condition until it was pointed out by the inspector. The problem was determined to be a defective detector prob c. Conclusions Several conditions that could have spread contamination or indicated a change in radiological conditions were identified during routine NRC plant walkdowns. Although no actual contamination issues resulted, the inspector considered that the conditions had likely existed long enough to have previously been identified by VY plant personne 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 September 7,1999. At this meeting, the purpose and scope of the inspection was reviewed, and the preliminary )
findings were presented. The licensee acknowledged the preliminary inspection findings, The inspector asked the licen.see whether any material examined during the inspection should be considered proprietary. No proprietary information was identifie '
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Attachment 1 LIST OF ACRONYMS USED BMO basis for maintaining operation CFR Code of Federal Regulation CRD control rod drive CS core spray l DBD design basis document !
ECCS emergency core cooling system ;
EDG emergency diesel generator EN event notification ER event report I LCO limiting condition for operation MIC - microbiologically induced corrosion NRC Nuclear Regulatory Commission RCIC reactor core isolation cooling RHR residual heat removal RHRSW residual heat removal service water RP radiation protection RRU reactor recirculation unit SW service water
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TM temporary modification TRU turbine recirculation unit TS technical specifications UFSAR Updated Final Safety Analysis Report W Vermont Yankee
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. Attachment 2 ITEMS OPENED, CLOSED, OR DISCUSSED l
OPENED IFl 99-06-01 Surveillance Test Value of RCIC Pump Discharge Pressure CLOSED None i
DISCUSSED None NON-CITED VIOLATIONS NCV 99-05-02 Inadequate Design Control for Torus-to-Reactor Building Vacuum Breaker Modification (page 8)
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