IR 05000271/1993026
| ML20059C825 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/22/1993 |
| From: | Eugene Kelly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059C680 | List: |
| References | |
| 50-271-93-26, NUDOCS 9401060119 | |
| Download: ML20059C825 (44) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
93-26 Docket No.
50-271 Licensee No.
DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road
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Brattleboro, VT 05301
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Facility:
Vermont Yankee. Nuclear Power Station Vernon, Vermont Inspection Period:
October 10 --November 27,1993 Inspectors:
Harold Eichenholz, Senior Resident Inspector Paul W. Harris, Resident inspector John T. Shedlosky, Project Engineer Daniel H. Dorman, Project Manager
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I.2/n/93 Approved by:
Eugene M. Kelly, Chief /
Date Reactor Projects Section 3A Scope:
Station activities inspected by the resident staff this period included Operations, Maintenance, Engineering, and Plant Support. - Assessment of a containment
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vacuum breaker QA audit finding and evaluation of radwaste system filtration efficiency were selected as initiatives this period. Backshift inspection (including weekend activities) amounting to 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> was performed on October 10,16,19, 22, and 26, and November 9, 21, and 23. Interviews and discussions were conducted with members of Vermont Yankee management and staff as necessary
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to support this inspection.
Findings:
An overall assessment of performance during this period is summarized in the
Executive Summary.
A violation for inadequate procedures involving
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maintenance (Section 3.1.2) and temporary material control (Section 4.2) was-
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identified. Vermont Yankee corrective actions in response to a previous violation i
(Inspection 93-14) failed to prevent the recurrence of unintentional modification of plant structures, systems, and components such as occurred this period.
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9401060119 931223 PDR ADOCX 05000271-G ppg
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EXECUTIVE SUMMARY Vermont Yankee Inspection Repod 93-26
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Operations Safe reactor startup and power ascension coming out of Refueling Outage XVII were observed.
The implementation of good control room shift briefs contributed toward shift personnel knowledge of plant conditions. However, details regarding abnormal plant conditions were not consistently documented in control room logs.
Maintenance and Testing The failure of a moisture separator emergency drain valve was identified by a licensee initiative
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to optimize feedwater heating, and a good questioning attitude by technicians. An environmental.
qualification (EQ) concern involving an unintended facility modification was caused by inadequate maintenance planning procedures. A thorough inspection of the primary pressure boundary during the reactor vessel hydrostatic test resulted in the identification of a through-wall leak in the reactor core isolation cooling (RCIC) steam supply line. The integrated emergency core cooling system (ECCS) test was conducted safely, including a comprehensive pre-evolution briefing, and good operator performance was observed.
Engineering P
A comprehensive evaluation supported the correction of a design deficiency associated with the service water system. Problems recurred regarding the lack of seismic evaluation and incorrect
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installation of scaffolding near safety-related systems.
Plant Support
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Comprehensive, probing discussions occurred during the Nuclear Safety Audit and Review i
Committee meeting for resolution of a QA audit finding involving a Technical Specification surveillance. Good industry event training was observed related to an industry event concerning
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the application of Furmanite sealant to an unisolable primary system valve. The evaluation of degraded radioactive waste ventilation system operation due to leakage around the filter seals
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lacked specificity and depth, and the licensce's Basis for Maintaining Operation process was not implemented in a timely manner.
Housekeeping near safety systems was good and has improved. Good response and judgement were demonstrated in response to fire alarms, and initiatives to improve Fire Brigade communications and access to locked areas were instituted.
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SUMMARY OF FACILITY ACTIVITIES Vermont Yankee Nuclear Power Station Refueling / Maintenance Outage XVII, and the return to power operation, were safely completed this inspection period. Surveillances were performed to assure satisfactory operation of the emergency core cooling systems, control rod drive system, primary plant integrity, and systems taken out of service for maintenance. Offgas activity levels at full rated power were approximately 7000 pCi/ml indicating good fuel integrity and accuracy of the pre-startup estimate. To provide additional assurance that Cycle XVII would remain free of defective fuel, reactor engineering recommendations involving conservative rod pulls and ramp rates were implemented.
An enforcement conference (Section 6.2) was held on November 23 regarding the September 1993 refueling events (NRC Inspection Report 93-81). On November 12, a management meeting was held to discuss Vermont Yankee (VY) initiatives (Section 6.3).
2.0 OPERATIONS (71707)
2.1 Operational Safety Verification The inspectors verified adequate stafGng, adherence to procedures and Technical Speci6 cation (TS) limiting conditions for operation (LCO), operability of protective systems, status of control room annunciators, and availability of emergency core cooling systems (ECCS). Plant tours confirmed that control panel indications accurately represented safety system line-ups. Safety tagouts properly isolated equipment for maintenance.
2.1.1 Reactor Startup and Power Ascension On October 23, VY commenced a normal reactor startup and ascended to full power operation.
Conduct in the control room was professional and team oriented.
Operators effectively communicated abnormal indications to supervisors and coordinated procedure reviews resulting in timely assessments. During startup, an initiative to control access to the control room contributed in a reduction of congestion and noise due to persons not directly responsible for safe plant operation. Operators were attentive to changing conditions and anticipated subsequent conditions, procedures, and surveillance requirements. Response to panel annunciations was
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prompt. Management oversight of activities was also evident.
Similar to previous outages, the conduct of the pre-startup Plant Operation Review Committee
(PORC) contributed to a comprehensive assessment of plant status to support a safe reactor startup. The PORC agenda included deferred maintenance, maintenance performed, surveillance results, as well as specific conditions associated with electrical switchgear, ECCS systems, and the emergency diesel generators.
Quality assurance reviews, radiological and chemistry assessments, and open items were also reviewed. No items were identified to preclude a safe power ascension.
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2.1.2 Control Room Shift Briefings On November 1, VY commenced conducting control room shift briefings in an effort to assure that information regarding the current status of safety systems and plant operations were understood by shift personnel. Plant procedure AP 0152, Rev.15, " Shift Turnover," was revised to state the detail and type of information presented and required attendees. Industry information and regulatory guidance from lessons learned from the Three Mile Island, Unit 1, accident (NUREGs 0585 and 0578) were included in the development of the briefing standard.
Teamwork was embodied in this initiative by requiring the attendance of Auxiliary Operators and radiation protection and chemistry personnel at all shift briefings. The briefings are conducted in the control room following watch relief, and have not reduced operator attention to plant conditions. Inspector observations this period confirmed that operators have a more comprehensive knowledge of plant conditions.
2.1.3 Lack of Detail in Control Room Logs On four separate occasions this inspection period, information regarding plant conditions were not fully described in control room logs. These occurrences included: (1) safety relief valve leakage following reactor startup from the refueling outage, (2) licensee concerns regarding problems associated with the primary containment nitrogen inerting system, (3) failure of the moisture separator emergency drain valve to fully close (Section 3.1.3), and (4) an engineering design change to the stack gas monitoring system.
In no case did the inspectors identify a lack of understanding of these conditions by either plant management or control room operators. Each item was adequately evaluated by the licensee to be of relatively low safety significance, and corrective actions were in place as required.
However, the lack of documentation of these occurrences in the control room log represented
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a weakness in logkeeping.
In an effort to improve logkeeping activities, the Operations Manager restated management expectations at the periodic Shift Supervisor meetings. In addition, information was provided in night orders to the operating staff regarding the need to improve logkeeping performance.
3.0 MAINTENANCE AND TESTING (62703,71500,62700)
3.1 Maintenance
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The inspectors observed selected maintenance on safety-related equipment to determine whether these activities were effectively conducted in accordance with VY TS, and administrative controls (Procedure AP-0021 and AP-4000) using approved procedures, safe tagout practices and appropriate industry codes and standards.
Interviews were conducted with the cognizant engineers and maintenance personnel and vendor equipment manuals were reviewe.
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3.1.1 Moisture Separator Emergency Drain Valve Failure
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i On October 29 the Operations Planning Coordinator (OPC) was conducting an initiative involving the balancing of the feedwater heater level controls. This is accomplished by adjusting the proportional band of the level controllers, which results in stabilizing the heater control valves, allows optimization of heater levels, and aides in minimizing heater erosion.
Subsequently, the OPC identified level oscillations on each of the four moisture separator (MS)
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level control valves. The control valves were observed to be continuously varying between 0-75 percent of valve stroke. A high level in any MS would result in a turbine trip and automatic
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The OPC's observations were provided to the I&C Department, which resulted in the initiation
of troubleshooting activities. The design of the MS level control system includes an emergency j
drain valve that automatically opens on a high level. It is the I&C Department's normal practice
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to first verify proper operation of the emergency dump valve prior to initiating any activity associated with the normal MS level control system. When a high MS level condition was l
simulated by I&C technicians, the "A" MS emergency drain valve " CV-23A) automatically i
opened, but did not reclose. Further efforts to reclose the valve te unsuccessful. An
engineering assessment identified no safety concerns for short term opersion of the plant that would be catred by the approximately 2-3 MWe loss that was occurring because steam was
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passing directly from the MS to the condenser.
The inspector reviewed design documentation and interviewed cognizant engineering and operations personnel to independently assess the potential for conditions adverse to plant or
personnel safety as a result of current operations. One potential concern involved accelerated erosion / corrosion due to the continuous high steam flow that would result from the stuck open valve.
The inspector learned that in 1992 the carbon steel (a material susceptible to steam erosion) piping downstream of the MS emergency dump valves was replaced with an erosion resistant chrome-moly material. A one-foot section of interface piping with the condenser, although originally made of schedule 80 carbon steel, was replaced with schedule 120 material,
which significantly increased the margins of safety. Finally, a review of 1992 emergency dump valve internal inspection results did not identify any significant valve body erosion. The j
inspector concluded that VY's erosion / corrosion program was effective in assuring that a good level of structural integrity was maintained to support short term abnormal operations of the subject system.
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J Vermont Yankee was unable to ascertain the rearon for the stuck open valve. Further efforts to resolve the condition would necessitate a plant shutdown, since the emergency drain line was not isolable. Subsequent to this inspection period, a plant shutdown was undertaken and a bolt was found lodged (and later removed) between the disc and seat. Regarding the abnormal control valve oscillations on the MS level control valves, the licensee's investigation determined that re-calibration of the level control system caused the observed performance. Vermont
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Yankee I&C personnel identified inconsistencies between calibrations of various components within the level control system, which resulted in changes to selected component calibrations.
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The I&C Department also identified that the original plant design for the MS normal level control valves had resulted in the installation of incorrectly sized control valves. This condition was initially encountered during original plant startup, but had been corrected by calibrating the control valve to be short-stroked (i.e., limiting valve movement to approximately 50% of normal travel). Recognition of the design issue allowed VY to implement interim repairs to return valve performance to the preoutage short-stroked condition.
3.1.2 Potential Environmental Qualification Concern On November 3,1993, the inspector observed that a 6 mil polyethylene plastic sheet was installed over the reactor building (RB) crane access hatch and questioned its effect on the environmental qualification (EQ) assumptions used to model high energy line breaks. Control room operators were unaware of this condition. Following identification of the concern, engineering guidance was requested. Within minutes the cognizant manager was apprised of the condition, and immediately directed removal of the plastic because its effect on the EQ of electrical equipment was unknown. Subsequently, VY determined that the plastic did not
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represent an EQ concern within the reactor building because the high energy line break environment would have displaced the plastic barrier, allowing natural circulation within the reactor building.
The plastic was installed on or about October 23 to control the spread of contamination during i
deconamination of the dryer separator pit on the RB refuel floor. The installation of the plastic was required by the work order (No. 93-9107). This maintenance is typically performed following refueling outages to establish good radiological conditions. The decontamination requires coordination between the radiation protection (RP) technicians and persons actually j
performing the work to prevent personnel contamination and control the spread of contamination.
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The inspector identified a number of concerns that occurred 6. ring the planning of this j
maintenance. First, maintenance planners and persons involved are not trained to identify this j
type of potential EQ concern created by maintenance. Therefore, it is not reasonable that
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persons involved should have recognized the EQ c meern in that it represented a potential change
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to the air flow patterns in the RB, vis-a-vis r. potential facility modification.
Second, maintenance procedures were not useful, in that, 6ey lacked effective instructions regarding EQ
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considerations of the type discussed in this event. Specifically, AP 0200, Rev.14, " Conduct of Maintenance," provides details on how to generate and plan a work order using the Maintenance and Plaiming Computer (MPAC) system; however, the fields presented on the work order did not key the planner into identifying that the installation of the plastic represents a potential EQ concern. This system does, however, provide good information for work planning and whether systems and components require EQ consideration. Further, Procedure AP 0021, j
Rev. 22, " Work Orders," does not adequately address the EQ concern discussed above. Third, J
communications regarding this maintenance were not entirely effective. Although VY credits the daily maintenance planning meeting as a mechanism to review maintenance prior to its performance, it appears that the issue was not discussed and, if it was, the meeting on November 3 was ineffective in addressing the potential EQ concer One corrective action was implemented prior to the end of the inspection period which consisted of increasing management awareness to EQ considerations. This was accomplished by a memorandum distributed to department managers and discussions at a daily Plant Manager meeting. No consideration was given to potential work control issues. The inspector concluded that insufficient instructions exist to reasonably conclude that recurrent issues will be precluded.
Technical Specification 6.5. A.I. requires, in part, that detailed procedures shall be written for maintenance. Also,10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures and Drawings," requires that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances. The VY EQ Program describes assumptions used to support analysis for the qualification of electrical equipment. The conditions identified above represent a failure to provide adequately detailed procedures to evaluate the effect of maintenance on EQ of safety related electrical equipment, and is considered one example of a violation (VIO 93-26-
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3.1.3 RCIC Weld Repair During system walkdowns performed at reactor pressure vessel (RPV) hydrostatic pressure, the licensee identified water leaking from an instrument tap in the Reactor Core Isolation Cooling (RCIC) steam supply line. The tap is used to measure steam line flow and is located at a pipe bend between the reactor pressure vessel and the inboard steam isolation valve. Work orders were initiated and repairs completed following reactor depressurization and lowering of reactor water level to support welding. The piping classification is Safety Class 2, USAS B31;1, as documented in the Final Safety Analysis Report (FSAR), Table 4.1.1.
Based on the licensee inspections, the pressure boundary defect appeared to be a 3/16 inch size porosity defect of a fillet weld connecting the instrument piping (3/4 inch, sch 160) to the steam line. Liquid penetrant and visual testing were performed by a Level 11 qualified nondestructive examiner during weld excavation and post-maintenance testing using Yankee Nuclear Services Division (YNSD) procedures YA-PE-1 and YA-VT-12, respectively. The NRC inspector reviewed the results and noted that the minimum examination and acceptance criteria were
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documented as required. Previous indications have not been observed. The identification of the leak demonstrated the thoroughness of the RPV hydrostatic test walkdowns.
3.2 Surveillance (61726)
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i The inspector reviewed procedures, witnessed testing in-progress, and reviewed completed surveillance record packages. The surveillances documented below were reviewed and were found effective with respect to meeting the safety objectives of the surveillance program. The i
inspector observed that the tests were performed by qualified and knowledgeable personnel, and j
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in accordance with VY TS and administrative controls (Procedure AP-4000), using approved
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procedures.
The following tests were found to be acceptable:
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OP 4114, Rev. 26, " Standby Liquid Control System Surveillance"
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OP 4360, Rev.19, "HPCI System Actuation Logic Functional / Calibration Test"
OP 43102, Rev. 2, "RCIC System Actuation Imgic Functional / Calibration Test"
3.2.1 Control Rod Scram Testing and Data Reduction The results of this testing in accordance with Procedure OP 4424 were consistent with the previous surveillance conducted April 15, 1993:
i 89 Rod Average - 0.313 seconds (TS limit: 0.358 seconds; previous: 0.324 seconds)
2X2 Array Average - 0.327 seconds (TS limit: 0.379 seconds, 4.51 % insertion;
previous: 0.340 seconds)
i 3.2.1 Integrated Emergency Core Cooling System Test On October 16, the inspector observed the conduct of the integrated emergency core coolant system (ECCS) test performed to verify the response of the emergency injection systems during a simulated coincident loss of coolant accident and unavailability of offsite power. The
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j surveillance also verified that the emergency diesel generators (EDGs) were capable of carrying design basis electrical load. The test was conducted in accordance with procedure OP 4100, Rev.19, " Integrated Emergency Core Cooling Tests."
Prior to the test, the inspector independently verified test prerequisites, performed a walkdown of control room panels, and verified that systems in a shutdown or maintenance condition did not affect the conduct of the test. The Operations Manager led a comprehensive pre-test brief with all individuals involved and covered responsibilities, exigent-conditions, and communications. The discussion regarding cetions in response to an actual emergency condition was an improvement from the last conduct of the surveillance. Control room operators were
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alert and cognizant of changing system status during the test.
A well-coordinated and professional response to a fire alarm outside of the operating EDG rooms was observed such that attention to plant and test conditions remained at an appropriate level. The test satisfied TS requirements and was conducted safely.
4.0 ENGINEERING (92701, 92702, 90712, 90713)
4.1 Temporary Modification of Service Water Pump Motor Cooling Line I
During this period, VY identified that the failure of a solenoid-operated isolation valve'(SOV)
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in the motor cooling line for the "C" residual heat removal service water (RHRSW) pump was j
caused by an inadequate design. The motor cooling line was originally designed to take supply
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l from the pump suction line at approximately 100 psig. However, during the mid-1970s, the l
cooling line supply was changed to the pump discharge at approximately 300 psig, to assure j
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operation. The 300 psig pressure exceeded the design rating (150 psig) of the cooling line, and ultimately caused plastic deformation of the SOV body and failure of one of its body-to-bonnet studs.
i Vermont Yankee identified this failure because water was leaking from the flanged surface of the solenoid valve. This modification was performed on all four RHRSW pumps, thereby l
subjecting all RHRSW motor cooling lines to pressures above design. Timely operability and failure assessments were performed, and a 4-hour notification to the NRC was made.
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As a result of the overpressure concern, the licensee modified the water supply to the RHRSW pump motor cooling support system. This modification orients the line to take suction from the
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sensing line of a pressure control valve on the RHRSW pump discharge, exposing the line to 100
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psig during both normal and alternate cooling modes of operation. On October 22, the Plant Operations Review Committee (PORC) reviewed the safety evaluation (SE) for the proposed modification and concluded that no unreviewed safety question existed. Prior to plant startup, the modification was appropriately implemented using AP 0020, " Temporary Modification.,"
The inspector observed the PORC and reviewed the SE; the evaluation was thorough and plant management involvement was evident.
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4.2 Seismic Evaluation of Scaffolding
NRC Inspection Repon 93-14 documented a concern regarding the lack of seismic evaluation of scaffolding installed in the vicinity of safety-related systems, structures, and components.
Vermont Yankee Procedure AP 0019, Rev. 6, " Control of Temporary and/or Portable Material,"
states that plant components must be protected to ensure safe operation of the facility, that damage to this equipment is avoidable providing the proper amount of care and attention to detail
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is exercised, and that the procedure is intended to provide guidance in preventing damage to
'i equipment. The procedure funher states that each situation is unique and must be carefully evaluated. Procedure AP 0019 does not specifically address seismic qualification; it does, however, state in Step B.3 that ladders must be tied off.
Information Notice 80-21, " Anchorage and Support of Safety-Related Electrical Equipment,"
informed licensees of concerns regarding the inadequate installation of temporary materials such j
that they may potentially dislodge, impact, and damage safety-related equipment during an j
earthquake. In 1988, VY controls of transient equipment were found to be adequate based on j
an NRC inspection using Region I Temporary Instruction No. 87-03, " Storage of Transient j
Equipment in Safety-Related Areas." During that review (NRC Inspection Report 88-06), no -
concerns were identified with AP 0019, " Control of Temporary Load on Piping, Equioment, and Structures."
On November 10, 1993, with the reactor at power, the inspector identified scaffolding deficiencies associated with the erection of temporary staging on the 213 foot level of the "B" ECCS comer room. The scaffolding was approximately 12 feet high, located between an RHR pump and the "B" core spray pump, and was not laterally restrained. In addition, the "B" core
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l spray pump was found to be within the fall zone of an access ladder that was not securely fastened to prevent falling. Poor control of temporary materials was also observed near the "A" reactor recirculation pump, in the control rod drive rebuild room, and at the north bank hydraulic control units. Those observations on November 8,1993, reflect a lack of rigorous
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control of temporary materials to prevent equipment damage or personnel injury as a result of
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an earthquake.
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l The inspector concludcd that the conditions identified on November 10, 1993, potentially
affected the operability of safety-related equipment located in the ECCS room during a seismic event, and that the installation was required to be evaluated for seismic conditions. Technical
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Specification 6.5. A.1 requires, in part, that detailed procedures shall be written for maintenance l
affecting safety. The conditions identified above represent a failure to both adhere to and implement detailed procedural requirements for seismic concerns, and are considered a violation (VIO 93-26-01). A previously identified unresolved item (URI 93-14-03) involving seismic i
evaluation of scaffolding is considered as part of this most recent finding.
4.3 Response to Violation 93-14-01
As documented in NRC Inspection Report 93-14, VY failed to identify the temporary
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modification (TM) of the reactor core isolation cooling (RCIC) alternate shutdown panel during the installation of temporary materials and scaffolding to support preventive maintenance in the vicinity of the panel. The TM consisted of a plexiglass sheet that covered and prevented access
to the controls of the alternate shutdown panel. The sheet was securely fastened with wire to the alternate shutdown panel support brackets.
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By 1ctter dated September 27, 1993, VY stated that the installation of the plexiglass was unacceptable and was due to their failure to properly implement plant procedure AP 0019, Rev.
19, " Control of Temporary and/or Portable Materials." Vermont Yankee stated that AP 0019 contained requirements to assure that temporary materials are not placed on or block access to safety-related equipment, and to assess the need for a TM request under AP 0020, " Temporary Modifications." Further, the licensee cited that inaccurate communications contributed to the deficient condition. Corrective actions included: (1) an interim program to conduct plant walkdowns with a system oriented checklist to identify adverse conditions; and (2) barrier installation to restrict access to vital plant equipment during the outage. Credit was also taken for the VY Observation Program and housekeeping tours. Long-term corrective actions included an intent to review lessons learned from outage housekeeping and TM installation and to revise
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procedures AP 0019 and AP 6024, Rev. 8, " Plant Housekeeping." The interim measures were to remain in place until the programs and procedures were appropriately revised.
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Although the inspector acknowledged these corrective measures, and agrees that the existing l
control of temporary materials should have prevented the condition identified on July 22,1993, the RCIC alternate shutdown panel was temporarily modified. However, the modification was not recognized by the implementing department supervisor, as required by AP 0019. Vermont Yankee stated that the plexiglass was installed "due to a lack of a system oriented review..."
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a however, no corrective actions were implemented to assure the timely and correct evaluation of temporary materials by the installing organization. The interim checklist cannot be credited as i
an effective corrective action, because it allows time to elapse between the installation of the temporary materials and the identification of the discrepancy. Further, Section 4.2 of this report documents NRC identification of another temporary material deficiency, indicating that the
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checklist and department training conducted on July 30,1993, were not effective.
Based on inspector review of the interim checklist and recurrence of another temporary material
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deficiency, the corrective actions identified by VY letter dated September 27,1993, do not
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provide sufficient assurance that AP 0019 or AP 0020 will be implemented effectively. This
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portion of the violation (VIO 93-14-01) therefore remains open.
5.0 PLANT SUPPORT (71707, 40500, 90712, 92702)
t 5.1 Industry Event Training During this inspection period, VY conducted training on a recent event that occurred at a
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pressurized water reactor on August 5,1993. This event involved multiple failure.s and
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subsequent repairs of a primary system valve during power operation. The repairs performed contributed to the ultimate failure of valve bonnet-to-body bolting and forced a plant shutdown due to elevated primary system leakage, and placed plant personnel in a potentially unsafe
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situation. Weaknesses associated with work implementation and control; safety review and
engineering support of maintenance; and quality verification of activities occurred during this
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event. All VY department managers attended the training and the Plant Manager discussed the weaknesses, lessons learned, the importance of assuring appropriate levels of management i
review, and the necessity to maintain a proper safety perspective for all plant activities.
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Effective training was also conducted with plant engineers, as verified by inspector interviews.
5.2 Degraded Radwaste Ventilation Exhaust System
Vermont Yankee Potential Reponable Occurrence Report (PRO) No.93-102 was written to
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address the current performance of the radwaste high efficiency particulate filters (HEPA) which
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i did not meet procedurally established filtering efficiency requirements. Subsequent management reviews concluded that the event was not reportable, and that maintenanx was required to l
i correct the degraded condition of the HEPA filters. An NRC Region I (NRC:RI) radiation
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effluents specialist independently confirmed that the dose and conditions identified were of low
safety significance (NRC Inspection Report 93-28). However, based on reviews and interviews i
performed, three weaknesses were identified.
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The degraded condition described in the PRO was not precise. Terminology used
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described the condition as " bypass Icakage" around the HEPAs; whereas the actual
condition represented a sealing degradation. The term " bypass" represents a condition in which ventilation flow is redirected around the HEPA filters; with " bypass" leakage occurring and the calculated site boundary dose exceeding the TS 3.8.1.1 action limit (as
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indicated on the PRO), the requirements of TS 6.7.C.2.b would apply. This TS requires a special 30 day report addressing the issue and corrective action, and was not fully considered by the licensee until questioned by the inspector. Subsequently, VY stated that due to conservatism associated with the initial calculation (a Method I technique)
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actual doses would be less than the TS action limit and a Method II technique would be performed to confirm this position. (See NRC Inspection 93-25 regarding Method I and
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II methodologies.)
The inspector noted that AP 0010, Rev. 24, " Occurrence l
Reports / Notifications and Reports Due," Appendix I, "Repons Due - Radiological," Item
A-9, reporting requirements of TS 6.7.C.2.b, is incomplete, in that, the reportability of gaseous effluents is not addressed.
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The evaluation of the condition was not comprehensive in that the Basis for Maintaining
Operation Guideline (BMO) was not implemented. The VY BMO Guideline is a process used to assure that conditions adverse to quality will be evaluated tojustify the continued operation of a system in a degraded mode (NRC Inspection Report 90-09). The degraded i
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condition described in the PRO represented an entry condition into the BMO process, because: (1) operation of the filter train in a " bypass" mode would be different than described in FSAR chapter 10.12.3.5 and system design specifications, (2) the radwaste ventilation system is a TS system (paragraphs 6.7.C and 3.8.I.1), and (3) the ventilation i
system was operated in a degraded condition due to leakage around the HEPA filter seals.
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The evaluation of the degraded condition for a TS required system was not timely. Any increase in offsite radioactive effluents that approaches or exceeds TS limits is a potentially significant issue requiring evaluation. In the case of the issue evaluated in the l
PRO, the degraded condition was first identified on June 29, 1993, failed its post-maintenance testing following temporary repair on July 8,1993, and then during the first week of September offsite doses apparently exceeded the monthly TS action limit.
Subsequent evaluations were not performed until recently questioned by the inspector.
VY management expectations were not met during the assessment of this degraded condition in j
that the BMO was not developed in a timely manner. At the end of this inspection period, the
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BMO and Method II calculation were not completed.
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5.3 Quality Assurance Audit Finding Disposition During the conduct of a QA audit in the area of Technical Specifications (No. 93-15), the Yankee Atomic Company Nuclear Services Division (YNSD) audit team determined that the performance of surveillance procedure OP 4115, Rev. 29, " Primary Containment Surveillance,"
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did not satisfy the surveillance requirement of TS 3.7.A.6.a.(2). This surveillance involves the suppression chamber-to-drywell vacuum breakers, and requires gravity closure of the vacuum breaker to within dimensional tolerances to prevent suppression chamber bypass during a loss of coolant accident. Vermont Yankee issued PRO No. 93-50 to identify and resolve the
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concerns involving the surveillance. Disposition of this concern required a multi-discipline l
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review by maintenance, engineering, and operations. Maintenance completed their preliminary
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assessment on the day the issue was identified and a comprehensive engineering review subsequently confirmed the maintenance assessment. The complexity of the issue demonstrated
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that the audit was of very good detail. The disposition of PRO 93-50 was detailed and of high quality. In addition, engineering recommendations intended to improve testing technique, valve maintenance, and an assessment of dimensional tolerances were implemented by plant management.
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5.4 Nuclear Safety Audit and Review Committee
On October 27,1993, comprehensive probing discussions were observed during the Nuclear Safety Audit and Review Committee (NS ARC). These discussions involved the corrective action program, disposition of NRC Bulletin 93-02, and human factors coacerns regarding refuel i
operations. The presentations made to NS ARC were of adequate detail and specific explanations
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resolved questions asked by NSARC. The discussions challenged VY positions regarding j
corrective actions and whether the root causes were actually understood by plant management.
In addition, generic issues were discussed regarding the VY's response to Bulletin 93 02 i
regarding their practice of " hand packing" fibrous insulation into the cracks of mirror insulation.
The NSARC was concerned about the quality and accuracy of their initial bulletin response.
Vermont Yankee management agreed that additional clarity was required and submitted a j
supplement to their initial response. An initiative to augment NSARC with additional experience
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and a perspective of senior management was accomplished by the addition of the Maine Yankee
Nuclear Power Plant Vice President and Manager of Operations to the committee.
i 5.5 Radiological Controls Inspectors routinely observed and reviewed radiological controls and practices during plant tours.
The inspectors observed that posting of contaminated, airborne radiation, radiation and high radiation areas were in accordance with administrative controls (AP-0500 series procedures) and
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plant instructions. Locked high radiation doors were properly maintained and equipment and i
personnel were properly surveyed prior to exit from the radiation control area. Plant workers were observed to be cognizant of posting requirements and maintained good radiological
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housekeeping.
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5.6 Security
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The inspector verified that security conditions met regulatory requirements and the VY Physical Security Plan. Physical security was inspected during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures.
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5.7 Firr Protection As discussed in Section 3.2.1, control room operator response was good for a fire alarm in the turbine loading bay located outside of the EDG rooms. Although the alarm was spurious, communications accurately described the situation, and the response by the Shift Engineer who is the Fire Brigade Leader (FBL) was timely. Th inspector performed a reactor building tour during the conduct of the integrated ECCS test and observed no deficiencies associated with fire loading in the EDG rooms or loading bay.
During observation of a Fire Brigade response to a fire alarm for the diesel fuel transfer pump house, the inspector observed that the Shift Engineer (SE) had neither the two-way radio nor the
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FBL's keys to access the pump house. The FBL intended to rely on the plant paging system to
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communicate with the control room and the Auxiliary Operator's keys. Procedurally, his
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guidance does not take into account that the SE periodically performs plant tours and does not require having the keys and radio.
The Operations manager provided clarification of management expectations that SEs are to be in possession of both a radio and keys when they are not in the vicinity of the control room. Further evaluations involving radio availability for the FBL are being conducted by Technical Programs Group.
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Good judgement was exercised by the SE to respond directly to the area of the fire alarm and
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l not return to the control room from the field. Also the inspector witnessed good response tactics by the FBL and fire brigade in assessing the alarm condition.
5.7 Ilousekeeping NRC inspections during Refueling / Maintenance Outage XVII verified that plant housekeeping in the ECCS corner rooms was acceptable. Temporary material staged for outage maintenance
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was controlled such that access to the ECCS components was maintained. Maintenance in progress on the residual heat removal systems did not affect the operability of the adjacent core
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spray system.
By memoranda dated October 11 and November 1,1993, VY assigned l
l housekeeping responsibilities by area, defined and clarified management housekeeping expectations, and set policy to require senior management supervision of subordinates to assure thM consistent housekeeping is maintained. On a number of occasions, the inspectors observed department managers touring the reactor building and discussing work control expectations with the maintenance crews. On October 18,1993, prior to plant startup, the inspector observed the Plant Manager and Technical Services Superintendent performing a detailed inspection of the feed pump, high pressure coolant injection, and ECCS comer rooms. Minor deficiencies were noted, corrected on the spot, and/or placed in the maintenance work order system.
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conditions adverse to system operability were noted. A sampling of the observations involving
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l housekeeping and material control were reviewed by the inspector and verified to be
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appropriately corrected.
Based on the housekeeping observed and corrective actions implemented, the inspector considered the prior identified NRC concerns resolved. The hcusekeeping portion of violation (VIO 93-14-01) is therefore closed.
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6.0 MANAGEMENT MEETINGS (30702)
6.1 Preliminary Inspection Findings
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i Meetings were held periodically with VY management during this inspection to discuss inspection findings. A summary of preliminary findings was also discussed at the conclusion of the inspection in an exit meeting held on December 3. No proprietary information was identified as being included in this report.
6.2 Enfortement Conference
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On November 23, 1993, an enforcement conference was held at the NRC:RI with VY representatives to discuss the Augmented Inspection Team (AIT) findings related to the fuel events (NRC Inspection Report 93-81). A list of meeting attendees and copics of overhead slides used in the VY presentation are contained in Attachments A and B.
t 6.3 Management Meetings A management meeting was held on November 12, 1993, at NRC:RI office with VY representatives.
Licensee initiatives and performance in the areas of self-assessment, management oversight, engineering, and refueling activities were discussed. A list of meeting attendees and copies of overhead slides used in the VY presentation are contained in Attachments I
C and D.
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ATTACHMENT A LIST OF ATTENDEES ENFORCEMENT CONFERENCE
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NOVEMBER 23,1993 NRC Attendees R. Cooper, Dimctor, Division of Reactor Projects (DRP)
C. Miller, Deputy Director, Division of Reactor Safety (DRS)
J. Linville, Chief, Projects Branch No. 3, DRP
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W. Butler, Project Director, Project Directorate I-3, Nuclear Reactor Regulation (NRR)
D. Dorman, Project Manager, Project Directorate I-3, NRR P. Harris, Resident Inspector, Vermont Yankee J. Beall, Team Leader, Engineering Branch, DRS
R. Barrett, Chief, Containment Systems and Severe Accident Branch, NRR A. Burritt, Operations Engineer, Operations Branch, DRS D. Holody, Enforcement Specialist, Office of Enforcement (OE)
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K. Smith, Regional Counsel E. Kelly, Chief, Reactor Projects Section No. 3A, DRP B. Whitacre, Reactor Engineer, DRP l
Licensee Attendees D. Reid, Senior Vice President, Operations R. Wanczyk, Plant Manager R. Pagodin, Operations Superintendent M. Mervine, Training Manager DJthe_I W. Sherman, State Nuclear Engineer, State of Vermont
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ATTACIIMENT B
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NRC/VY ENFORCEMENT CONFERENCE SLIDES l
NOVEMBER 23,1993
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INCIDENT OVERVIEW / SAFETY SIGNIFICANCE
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Prior Expectations
- Refueling-SALP
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e IndustryNY Experience
- Hardware improvements
- CEP Project
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- Training
e Management Expectations
- Procedures
- Briefings
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- Outage Safety initiatives
- Licensed Operator Training
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e Refueling incidents
- 9/3/93
- 9/9/93
. Lowering of Bundle
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. Management Investigation
. Procedure Use
. NSARC Review
. Safety Significance t
e Summary
- Cause Operator Errors
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Weakness in Management Oversight NEIM)MYANIGE i-NUcLEA OWER CORPORATION
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CYCLE 93.3 TOPICS (Focused on S/D Safety)
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Reactor S/D scenario in simulator.
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Power escalation scenario in simulator with
miscalibrated nuclear instruments.
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- Selected SOER's..
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RPV instrument leg problems.
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Loss of S/D cooling scenario in simulator.
- i Plant modifications and procedure changes.
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Refueling procedure review.
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I NEIUdBMTYANIEE i-
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NUCLEAH POWER CORPORATION
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LONG TERM CORRECTIVE ACTIONS
i e Engineering review oflesson plans involving plant modifications.
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i e Benchmarking of refueling training vs.
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other BWR's
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- Development of a more extensive fuel handling c
course (including OJT) to be used for both
- initial and requalification training.
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- Self-checking training for appropriate personnel.
- Review other infrequently performed safety j
significant tasks to ensure periodic t
requalification training is adequate to maintain
operator skills.
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NEIN$ POWER CORPORATIONMTYANIG li-NUCLEA
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PRIOR TO REFUELING
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Management Focus on Operator Performance
- Command and Control
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- Communications
- Procedure Adherence Hardware Upgrades e
- Eliminate Refueling Distractions
- Based on VY and Industry Experience
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- Improved Level of Control
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Outage Staffing issues e
- Activated 2 additional SRO Licenses
- Four Shift Rotation
- Specific Refuel Schedules, Limited Hours e Shift Supervisor /SRO Meetings
- Focus on Outage Safety
- Observation Program /Self Assessments
- Command and Control e Management Oversight of Refueling
- Provided by Dedicated SRO
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FOLLOWING 9/3 REFUELINGINCIDENT
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- Additional Hardware Modifications
- Interlock to Preclude Raising Without Grapple Closed
- Added Demarcation to Bridge Controls i
introduced Self Checking e
- Memo from Plant Manager Addresses:
I STAR Procedure Adherence Attention to Detail
- Questioning Attitude
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Developed Specific Training o
- Command and Control
- Communications
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- Procedure Adherence
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Management Oversight
- Provided by QSG and OPS Management
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- included OJT
- Included initial Fuel Movement for l
Each Crew
l NEl&$ANTYANIEE i-NUCLE POWER CORPORATION l
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l FOLLOWING 9/9 REFUELINGINCIDENT e Additional Hardware Modifications
- Reversed Direction of Joystick
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- Changed Color of Grapple Closed Indicator
- Removed Unofficial Operator Aids
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Complete Review / Revision of Refueling e
Procedure
- Required All Initial Movements to be Slow, Deliberate i
- Clarified Responsibilities
- Required Pre-Job Briefing
- Procedure Required on the Refuel Floor
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Management Expectations Reinforced o
- Plant Manager Directive on Procedure
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Adherence
- Formal Training on Procedure Changes
- Formal Training on Self-Checking
- Plant Manager Met with Refuel Personnel
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Management Oversight s
- Full-time QSG Surveillance
- Full-time OPS Management Oversight
- Part-time Senior Plant Management Oversight NERM$ POWER CORPORATIONNTYANIG i-:
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LONG-TERM CORRECTIVE ACTION l
Hardware issues
- Continue with Phase 2 of Bridge
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Modifications
- Human Factors Review
- Operators input
- Permanent Addition of Grapple Interlock
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Procedures
- Benchmark Refueling Procedure
- Re-evaluate Fuel Move Sequence
- Assess Responsibilities, Authorities, Clarity
- Assess Additional Human Factors
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Review of Generic Implications NERI < *4NTYANIGE i-1 NUCLEA POWER CORPORATION
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GENERICISSUES/CORRECTIVEACTIONS
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o Design Change improvements
- Human Factors improvements
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Training
- Engineering / Training Interface
- Self-Checking
- Review of Safety-Significant/ Infrequent Tasks
e Management
- Self-Assessment of Corrective Actions (QA, VY)
- Procedure Use/ Compliance
- QA Assessment of Safety-Significant/ Infrequent Tasks
- Lessons Learned Review - Other Departments
- Review of Safety-Significant/ Infrequent Tasks
e Other
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- Observation Program
- Self-Assessment Program
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- OP 4179 Assessment
- Management Expectations
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Daily Meetings SS Briefings
- NSARC Involvement in Audit Process
- Management Observations
- Simulator i
NERM&WYANKEE i-
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SAFETYSIGNIFICANCE
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Design Basis
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- Radiological
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- Systems Management Oversight
- Design Changes
- Training
- Procedures
- Command and Control
- Management Expectations
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NERM$ POWER CORPORATIONNTYANIGE r-NUCLEA
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SUMMARY
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Identification and Reporting
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- Halted Refueling Activities
- Prompt NRC Notification
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Corrective Action
- Timely
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- Extensive
- Comprehensive
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Past Performance
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-SALP
- Previous
- OSTI
- Recent VY Initiatives
Prior Opportunity
- Audits / Surveillance's/ Inspections
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- Industry
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Occurrences / Duration
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ATTACIIMENT C
LIST OF ATTENDEES CORRECTIVE' ACTIONS
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NOVEMBER 12,1993
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NRC Attendees:
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T. Martin, Regional Administrator R. Cooper, Director, Division of Reactor Projects (DRP)
W. Hodges, Director, Division of Reactor Safety (DRS)
S. Shankman, Deputy Director, Division of Radiation Safety and Safeguards (DRSS)
J. Linville, Chief, Projects Branch No. 3, DRP
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E. Kelly, Chief, Reactor Projects Section 3A, DRP H. Eichenholz, Senior Resident Inspector, Vermont Yankee
W. Butler, Director, Project Directorate I-3, Office of Nuclear Reactor Regulation (NRR)
Licensee Attendees:
J. Weigand, President and CEO D. Reid, Vice President, Operations
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J. Pelletier, Vice President, Engineering R. Wanczyk, Plant Manager l
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ATTACIIMENT D NRC/VY MANAGEMENT MEETING SLIDES NOVEMBER 12, 1993 i
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NRC/VY MANAGEMENT MEETING November 12,1993 AGENDA introduction J.G. Weigand Self Assessment / Oversight Initiatives Corporate D.A. Reid/J.P. Pelletier
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QA W.K. Peterson
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Plant R.J. Wanczyk Refueling Summary R.J. Wanczyk Conclusion J.G. Weigand VY ATTENDEES J.G. Weigand - President and CEO D.A. Reid - Vice President, Operations J.P. Pelletier - Vice President, Engineering R.J. Wanczyk - Plant Manager W.K. Peterson - Manager of Quality Assessment
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PERFORMANCE REVIEW COMMITTEE
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Expectations / Philosophy of Management Needs improvement
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Enhanced Management Training
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PERFORMANCE REVIEW COMMITTEE
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Precursors Are Not Always Recognized
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EVIDENCE Power Monitor.
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Core Spray Suction Strainers
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Turbine inspections
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AO - Core Spray DP Switch
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Seismic-Electrical Buses
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Startup Transformer Ground
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Management Outage Critique
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ENGINEERING INITIATIVES 50.59 ENHANCEMENTS
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DESIGN PROCESS IMPROVEMENTS
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EVALUATION IMPROVEMENTS j
i NEW LOCA AND CONTAINMENT
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ANALYSES y
SERVICE WATER SELF ASSESSMENT
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COMMUNICATIONS-INTEGRATED AUDITS AND
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NSARCINVOLVEMENT
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INCREASED MANAGEMENT
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