IR 05000271/1993007
| ML20044F823 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 05/18/1993 |
| From: | Lusher J, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20044F819 | List: |
| References | |
| 50-271-93-07, 50-271-93-7, NUDOCS 9306010090 | |
| Download: ML20044F823 (17) | |
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l U. S. Nuclear Regulatory Commission Region I Docket / Report:
50-271/93-07 License: DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Brattleboro, Vermont 05301-0169 Facility Name:
Vermont Nuclear Power Station Inspection:
April 26-29,1993 Inspection At:
Brattleboro and Vernon, Vermont Inspectors:
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J. Lusher, EgergerMy Preparedness (EP) Specialist date E. McCabe, Chief, Emergency Preparedness Section A. Mohseni, NRR:EPB EP Specialist H. Eichenholz, Senior Resident Inspector, Vermont Yankee P. Harris, Resident Inspector, Vermont Yankee C. Carroll, NRC Consultant (Sonalysts)
P. Reagan, NRC Consultant (COMEX)
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sow b WW Approved:
/E. M merfency Preparedness Section date
.Jis%e, Chie ion of Rad $
Safety and Safeguards Areas Inspected The licensee's annual, full-participation emergency preparedness exercise.
Results A generally good licensee emergency response was evident. No exercise weaknesses were identified. The scenario aggressively challenged the licensee response in the Operations Support Center (OSC), where team briefings, team dispatch, and procedure adherence were much improved over the previous exercise. But, there were some problems in this area, which will be held open pending further demonstration of improved OSC performance during regularly scheduled exercises. Emergency Response Organization (ERO) mobilization timeliness (95 minutes versus a 60 minute goal) and shift operator performance were identified as significant areas for potential improvement. No violations of regulatory requirements were identified.
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9306010090 9aosts
PDR ADDCK 05000271
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Table of Contents
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1.0 Persons Contact ed....................................... 3
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2.0 Emergency Ex ercise..................................... 3 2.1 Scenario Planning................................... 3 i
2.2 Exercise Scenario................................... 4
2.3 Activities Observed.................................. 5 2.4 Exercise Finding Classifications
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2.5 General Exercise Observations.......................... 6 2.6 Emergency Response Facility (ERF) Observations
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2.6.1 Exercise Data Adequacy and Communication to Players........
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2.6.2 Activation of Emergency Response Facilities (ERFs).........
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2.6.3 Simulator Control Room (SCR) Observations..............
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2.6.3.1 Shift Discipline and Professionalism (8)
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2.6.4 Technical Support Center (TSC)...................... 9 2.6.5 Operations Support Center (OSC)....................
2.6.5.1 Consideration of Plant Conditions and Documentation
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of Information (13)
2.6.6 Emergency Operations Facility (EOF)
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2.6.6.1 Core Damage Assessment (15);
2.6.6.2 Overall Event Response Timing (15)
2.7 Media Center.....................................
3.0 Licensee action on previously identified items
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4.0 Licensee critique.......................................
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5.0 Exit M eeting..........................................
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DETAILS
I 1.0 Persons Contacted The following individuals were contacted during the inspection and attendedd the exit meeting.
G. Bristol, Community Relations Coordinator
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T. Burda, SWEC, Emergency Planner L. Cantrell, Shift Supervisor
i A. Chesley, Technical Support Training Supervisor B. Finn, Senior Operations Training Instructor J. Herron, Technical Services Superintendent S. Jefferson, Assistant to Plant Manager, Exercise Coordinator
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M. Krider, Simulator Supervisor D. McElwee, Liaison Engineer A. Mears, Internal Auditor R. Pagodin, Operations Superintendent
J. Pelletier, Vice President, Engineering
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D. Porter, Engineering Projects Superintendent E. Porter, Emergency Planning Coordinator E. Salomon, Senior Engineer, Yankee Atomic M. Schneider, Mwager of Communications
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G. Sherer, TWC, be.;urity Controller J. Sinclair, Director, External Affairs R. Sojka, Operations Support Manager i
R. Wanczyk, Plant Manager The inspectors also contacted other licensee personnel.
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2.0 Emergency Exercise t
A full-participation emergency exercise was conducted at the Vermont Nuclear Power Station on April 27,1993 from 0200 to 1000.
2.1 Scenario Planning Exercise objectives were submitted to Nw Region I on Dece-mber 18,1992. The completed scenario package was submitted to the NRC on January 18,1993. Region I reviewers discussed scenario improvements with the licensee's emergency preparedness staff on March 3,1993. The
scenario adequately tested the major portions of the Emergency Plan and Implementing Procedures and demonstrated areas previously identified for corrective action.
On April 26, 1993, NRC observers attended a licensee briefing on the revised scenario. The-
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licensee stated that certain emergency response activities would be simulated and that controllers
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would intercede in exercise activities to prevent disrupting plant activities.
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2.2 Exercise Scenario The submitted scenario included the following simulated events:
Initial conditions: the plant was operating at approximately 100% power for an extended period and was nearing the end of the operating cycle.
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Both start-up transformers were lost due to failure in the start-up transformer lockout
relay. (An Unusual Event)
An earthquake, sensed on-site, resulted in damage to non-safety-related equipment.
e Feeder breaker DC-3A trip caused a loss of control room panel (CRP) annunciators on panels 9-3, 9-4, 9-5, 9-6 and 9-8. (An Alert)
Power to feeder breaker DC-3A was restored.
- The main turbine experienced excessive bearing vibration, causing a turbine trip.
e The reactor failed to scram (trip).
- Loss of normal power occurred. (A Site Area Emergency)
The service water flow control valve to the "B" diesel generator failed closed and resulted in a high temperature trip of the "B" diesel generator, e
Drywell vacuum breakers opened (one failed open) causing a direct flow path from the drywell to the torus atmosphere.
Repairs to the start-up transformer lockout relay were completed.
e Turbine Building ventilation was restored.
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All control rods were inserted.
- A small steam leak into the primary containment caused primary containment pressure to approach 5 psig.
- The hardened vent rupture disk began to leak, providing a direct path from the torus to the plant vent stack.
The plant vent stack indicated a release of radioactivity. (A General Emergency)
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Efforts continued to restore power to motor control center MCC-7, repair flow control
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e valve FCV-28B and isolate the hardened vent.
e The MCC-7 feeder breaker was repaired.
e Power was restored to the hardened vent isolation valve and the valve was closed.
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e The release was terminated.
- Off-site monitoring and dose assessment activities continued.
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e The exercise was terminated.
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2.3 Activities Observed The inspectors observed the activation and augmentation of the Emergency Response Facilities and the actions of the Emergency Response Orga:.:zation (ERO), including:
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Selection and use of control room procedures.
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Detection, classification, and assessment of scenario events.
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Notification of licensee personnel and off-site agencies.
Accident analysis and mitigation.
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Accountability of personnel.
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Provisions for in-plant radiation protection.
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Communications /information flow, and record keeping.
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Direction and coordination of the emergency response.
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Assessment and projection of radiological doses.
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Protective action recommendations (PARS).
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Provisions for communicating information to the public.
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The licensee's post-exercise critique.
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2.4 Exercise Finding Clnssifications Inspection findings were ca sified as follows:
f Exercice Strencth_; a strong positive indicator of the licensee's ability to cope with abnormal conditions and implement the emergency plan.
Exercise Weakness: less than effective Emergency Plan implementation which did not, alone,
constitute overall response inadequacy.
Area for Potential Imorovement: an aspect which did not significantly detract from the licensee's response, but which merits licensee evaluation for corrective action.
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2.5 General Exercise Observations Activation and utilizadon of the Emergency Response Organization (ERO) and Emergency Response Facilities (ERFs) were generally consistent with the Emergency Plan and Emergency Plan Implementing Procedures (EPIPs).
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2.6 Emergency Response Facility (ERF) Observations No exercise strengths or weaknesses were observed.
The following areas for potential improvement were noted.
i 2.6.1 Exercise Data Adequacy and Communication to Players
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When the simulator failed, as further discussed in section 2.6.3, the controllers had to change the computer-generated data on the spot so that it made sense to the operators.
In addition, the controllers could not generate some of the data that was requested by plant personnel, primarily because they were resource limited.
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At 0345, there was a simulated earthquake. At 0358, a facility controller handed the
Shift Supervisor a previously prepared message informing him that the plant I&C technician reported that the seismic computer indicated that the operating basis earthquake (OBE) was not exceeded. For a real eanhquake, this information would not have been available so quickly. When the facility controller was questioned by the NRC evaluators about this repon, he stated that he had made a mistake. The facility controller then informed the Shift Supervisor to disregard the OBE report. However, the Technical Support Center (TSC) and the Commonwealth of Massachusetts had already been relayed the message. This caused unnecessary confusion.
In most instances, Operational Support Center (OSC) controllers followed an appropriate
protocol to provide players with information. In some instances, however, controllers either voluntarily provided information or asked leading questions. Examples included the following.
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At approximately 0435, while doing a reactor building tour, the Auxiliary Operator (AO) received a message from the controller to the effect that a significant leak existed due to a flange leak in the cooling water outlet line from RRU-10, and that water was spraying in the vicinity of the 280' elevation. At that time, there was known to be excessive leakage into the Northeast floor drain sump. The AO appropriately relayed this information to the control room, then proceeded to continue with his reactor building rounds. The controller then asked if the AO intended to isolate the leak. The AO considered this question in concen with control room personnel and the decision was then made to isolate the leak before going elsewhere.
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At approximately 0453, a controller voluntarily offered information without being queried by the AO. This controller pointed out all the obvious problems during the reactor building walk-down and stated that there were no other problems, such as loose anchor bolts, leaks in the lower levels of the radwaste building, or leaks in the instrument air system.
2.6.2 Activation of Emergency Response Facilities (ERFs)
NUREG-0737, Supplement 1, specifies a goal of activation of the ERFs within about sixty minutes of declaration of an emergency (Alert or higher). During this exercise, the ERFs were activated as follows.
The Technical Support Center (TSC) was activated in about 85 minutes after the Alert-
was declared.
The Operations Support Center (OSC) was activated, in about 82 minutes after
declaration of the Alert. (There were indications the OSC was functioning sooner. For example, about 40 minutes after Alert declaration, the OSC asked to be informed about inspections being done per OP-3127.)
The Emergency Operations Facility (EOF) was activated about 85 minutes after the Alert
was declared.
2.6.3 Simulator Control Room (SCR) Observations Some problems resulted from simulator failure during the exercise. For example:
At 0430, there was an unplanned failure to scram (trip) and a loss of normal power in
the SCR. The crew was taking required actions when ti; simulator operators stopped J.: scenario and reset the computer. This problem was resolved in a few minutes and did not cause significant confusion because it occurred prior to staffing most outside organizations and because the malfur.ctinn was a short one.
At 0600, the coatrol room operators were attempting to reduce reactor power due to a
decreasing condenser vacuum. As they drove control rods into the core, reactor power increased. This was caught by the controllers and did not adversely affect the exercise scenario.
At 0635, the planned failure to scram was initiated, but the high pressure coolant
Injection (HPCI) and the reactor core isolation cooling (RCIC) syste-ms did not respond properly due to a simulator malfunction. Several minutes later, there was total loss of.
the simulator. The facility controllers were quick in providing verbal cues to control room personnel based on previously generated computer data. Their efforts allowed the exercise to continue.
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While the simulator problems were an adversity, NRC review concluded that continued exercise j
conduct was appropriate and adequately planned for.
Considering the circumstances, the following SCR findings were made.
There were no SCR exercise strengths or weaknesses.
In each instance where the Shift Supervisor (SS) was responsible for determining an emergency action level (EAL), he correctly assessed the plant's condition and made the proper classification.
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The control room communicator not only notified the NRC and various State agencies within the prescribed time, but also aggressively pursued those State agencies who did not call the control room back to confirm that they had been notified.
2.6.3.1 Shift Discipline and Professionalism i
The control room operators did not always execute their emergency operating procedures (EOPs)
in a disciplined and methodical manner. Specific examples follow.
The plant was experiencing an anticipated transient without scram (ATWS) with a failure
of both HPCI and RCIC. The Shift Supervisor ordered a second control rod drive (CRD) pump started to increase reactor water level. He took this action without first consulting the EOP for Level Control. After the second CRD pump was started, he checked the EOP and confirmed that he had taken the correct action.
The control room operators were executing OE-3106, Radioactivity Release Control
Procedure, which directed them to emergency depressurize the reactor in accordance with OE-3102. Without referring to OE-3102, the Shift Supervisor directed the reactor operator to open all safety-relief valves immediately. When this action was taken, none of the residual heat removal (RHR) pumps were in the pull-to-lock position as required by procedure. It was not until the reactor was actually blowing down that the operators prevented the RHR pumps from injecting.
E Shift discipline and professionalism degraded significantly after the simulator malfunctioned and the exercise controllers resorted to hard copy data to simulate reactor plant status. Some control room operators became detached from the scenario, and became involved in activities which did not contribute to the assessment, evaluation, and diagnosis of the plant release. Consequently, the operating crew derived little benefit from the team approach to problem solving. Specific
indicators of lapses in professionalism follow.
Once the simulator failed, the simulator operations crew did not aggressively utilize their
EOP flow-charts. The inability of facility controllers to provide information at a sufficient rate to fully occupy the control room personnel was a contributing factor. But,
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it was also noted that the operators did not use the EOPs to evaluate the data which were made available to them by the controllers, in that use of the EOP flow-charts after the
loss of the simulator was sporadic.
While a radioactive release was in progress, one operator made and ate a sandwich. This
same operator left the control room to get coffee without informing anyone else on the crew on at least one occasion, returning a few minutes later. During this operator's absence, the Shift Supervisor was attempting to locate a print in an effort to determine
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the location of the leak. At no time during the post-ATWS scenario did the NRC evaluator observe this individual operator attempt to assist other crew members.with the
evaluation and diagnosis of the radioactive release.
The Shift Supervisor found it necessary to remind the crew several times that they had
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to maintain their discipline to focus their attention on the plant status.
t On several occasions, the Shift Supervisor left the control room to smoke a cigarette with
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no formal turnover to the Assistant Control Room Operator (ACRO).
Overall, although data discrepancies made the simulator crew's job more difficult, the above examples showed substantial room for operator performance improvement. This was classified as an area for potential improvement (IFI 50-271/93-07-01).
2.6.4 Technical Support Center (TSC)
No exercise strengths or weaknesses were observed. However, the following actions were done
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j The Duty Call Officer promptly assessed plant conditions, effectively coordinated in-plant
response, and assumed accident management responsibilities from the Shift
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Supervisor / Plant Emergency Director. The accountability of perconnel and establishment i
of emergency communication links were timely. Status boards and the TSC log were
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accurate and used effectively, but meteorological data were not readily apparent. Upon
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loss of the plant simulator, the TSC staff demonstrated excellent professionalism by
continually evaluating plant parameters and determining trends and changes in plant
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status.
The TSC staff demonstrated good ability to perform accident assessment and event
classification. Discussions between the TSC and EOF regarding operational data and procedural clarifications during the ATWS and a reactor water level excursion led to
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prompt and accurate Site Area and General Emergency declarations. Reactor conditions l
were continuously assessed through the periodic sampling of reactor coolant and stack gas samples and the use of the plant process computer.
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I The TSC Coordinator (TSCC) effectively prioritized tasks based on the resulting
contribution to plant safety. Responsibilities were clear and effectively delegated. Staff
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meetings and Technical Support Center Coordinator / Site Recovery Manager briefs were frequent, contributed to event analysis, and clarified plant status and corrective actions implemented. The TSC reviewed the procedures implemented by the control room and provided timely and correct direction to assure the completion of safety-significant activities.
In general, thorough evaluations of plant conditions contributed to accurate event analysis e
and decision making. The TSC staff utilized the Engineering Support Group to evaluate seismic event equipment damage and equipment failure mechanisms, to inspect plant equipment, and to assess primary containment response during reactor vessel emergency depressurization. However, the safety significance of a stuck open suppression-chamber-to-drywell vacuum breaker to post-accident mitigation was not fully assessed. A delay in the determination of the release pathway resulted because reactor building temperatures and radiation levels were not effectively evaluated.
2.6.5 Operations Support Center (OSC)
i Vermont Yankee (VY) had established an ambitious schedule of mini-scenarios to improve in this area. Revisions were made to procedures (AP-0021, Work Orders; OP-3507, Emergency Radiation Exposure Control; and the Emergency Plan Implementing Procedures) to improve the
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controls and performance of the emergency on-site assistance teams. There were a number of notable positive observations involving: (1) plant Auxiliary Operator (AO) knowledge and use
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of Emergency Operating Procedure OP-3107, Appendix I, for local firing of a squib valve; (2)
the diligence of personnel in the OSC, and those assigned to field activities, in demonstrating
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response activities; (3) the implementation of proper radiological controls necessary to support OSC activities; and (4) the Instrumentation & Control (I&C) engineer's performance in i
demonstrating the acquisition of seismic monitoring data.
Several " mini-scenarios" occurred during the exercise. Inspectors observed these, including
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inspection of reactor building equipment, start-up transformer troubleshooting, retrieval of seismic monitoring data, repair of motor control center MCC-7, and the local firing of a standby
liquid control (SLC) system explosive squib valve.
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l To minimize activity simulation, various mockups allowed work crews to actually do some i
troubleshooting and repairing of components. Notabic among these were the mockup of the i
CRP-9-22 86 ST lockout relay (associated with start-up transformer trout % shooting), the mockup for the motor-operated torus vent valve, and the mockup of MCC-6 breaker 6T7 (associated with restoration of power to MCC-7). These mockups created more realistic mini-scenarios and facilitated demonstration of response personnel competence.
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The licensee had restrained the breaker 6T7 mockup, mounted on a wheeled cart, by using rope to tie it to MCC-6 Transformer T-6 (non-safety-related) cooling coils. When questioned by the inspector, the OSC controller promptly had the Breaker 6T7 mockup moved to an appropriate, p
nearby location.
Control room personnel dispatched an AO to check for cquipment damage and leakage caused l
by the postulated seismic event. The AO diligently and systematically checked all reactor building areas that were readily accessible. This AO periodically communicated his findings to the control room and obtained instructions for actions to isolate simulated leakage in two instances. Upon completing the reactor building tour, the AO reviewed OP-3127, Natural Phenomena, Appendix A, Seismic Damage Indicator Walk-down Check Sheet. OP-3127 required, for example, checking the anchor bolts for instrument racks RK-12A and RK-12B on the 318' elevation, but this was not done during the AO's reactor building walk-down. This discrepancy was an example indication that procedure adherence needs furtherlicensee and NRC attention. (See Report detail 3.0.)
The licensee's 1993 Plume Pathway Exercise Manual (PPEM), Section 3.1.C.1, stated (in part)
that: " simulation of response activities will occur only when actions are outside of the defined
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mini-scenarios." PPEM Section 3.1.B allowed controllers to "makejudgement decisions to keep the action going in accordance with the scenario time line." In some instances, OSC controllers
directed that an activity be simulated for the purpose of k.eeping the overall drill scenario on schedule. Related examples noted during this exercise follow.
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Completion of replacement part installation, post-work tests, and clearance of tags in Maintenance Request MR-93-002 and MR-93-003 was simulated at about 0723 and 1053, u
respectively.
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Simulations were done for MR-93-002. The OSC controller asked the affected electrical j
maintenance and operations personnel to describe the tasks to be done and generally applicable requirements for processing the MR to completion.
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Simulation of lockout relay work completion was allowed to expedite restoration of
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electrical power to benefit the overall exercise time line.
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To expedite restoration of MCC-7, the OSC controller acted as the electrical maintenance supervisor, and was authorized to do so by the OSC lead controller.
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An OSC controller unrealistically simulated that, within two minutes, an AO could retrieve a long ladder (plus gloves and other safety equipment, as needed) from
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somewhere in the reactor building, return to the 345' Elevation, and then close valve 146A, the steam supply isolation valve for start-up heater SUH-20.
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(6)
About 0615, an OSC controller allowed simulation of the prompt withdrawal of a replacement part (a relay) from the warehouse. At that time, warehouse personnel were not on duty and the work crew did not physically verify the part to be in stock, but they did use the Maintenance Planning and Control (MPAC) system to check warehouse inventory. When questioned by the inspector, the OSC controller asked the lead OSC controller ifit was necessary to more closely simulate withdrawal of the above described part from the warehouse. Based on the OSC lead controller's direction, the OSC controller then had the work crew adequately demonstrate the desired replacement relay could be obtained from the warehouse.
The number and nature of such examples indicated that, although a much improved approach to exercise realism was evident during this exercise, further minimization of activity simulation could be beneficial to training of the OSC staff. In consideration of the considerable licensee effort and progress on minimizing simulation, however, the inspector had no further questions on this matter.
MR-93-003 included simulation of breaker 6T7 repairs to expedite restoration of MCC-7 and closure of the torus vent isolation valve. In this case, the scenario allowed a temporary modification involving installation of a breaker without its two closing fuses. It was not clear whether that temporary modification was done per AP-0021, work order or 10CFR50.54(x).
Inasmuch as the actions involved were assessed as reasonable and necessary, the inspector had no further questions on this matter.
The inspectors also observed that, although the OSC was congested, there were no consequent exercise performance inadequacies. Good OSC performance examples included the following.
A status board outside the OSC Coordinator's (OSCC's) office listed principal activities
and work priorities.
Form VYOPF-3507.02, Emergency OSC Team Briefing / Debriefing, was used to document OSC team work assignments. Completed VYOPF-3507.02 forms were posted on a bulletin board in a hallway in the OSC area.
- When the OSCC left the OSC to attend meetings, he announced that the Assistant OSCC was in charge. That clearly communicated the temporary change in OSC command structure that had taken place.
Diligent procedure compliance was evident during the racking down of 4Kv Bus 2
breaker 27 and the racking out of MCC-7 breaker 77. Also, from about 0924 to 0941, an AO used Procedures OP-2142 and OP-2143, respectively, to effectively complete these breaker tasks.
The simulated accomplishment of the local firing of a SLC squib valve demonstrated
effective procedure utilization. (It was noted, however, that OP-3107, Appendix I did not contain information that clearly identified the SLC squib valve polarizing pin
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location.)
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13 Overall, no OSC exercise strengths or weaknesses were observed. Nonetheless, performance of the OSC was notably improved overall. Diligent procedure compliance, limited simulation J
of work activities, and appropriate communication between players and controllers generally l
characterized OSC performance. The following area for potential improvement was also identified.
2.6.5.1 Consideration of Plant Conditions and Documentation of Information Evaluation of the following revealed apparent problems in considering physical plant conditions and recording required information on Forms VYOPF-3507.01 and 02. These were identified as an area for potential improvement (IFI 50-271/93-07-02).
During the loss of normal power (LNP), the team assigned to perform local firing of the
SLC squib valve on the reactor building 318' elevation used the elevator, which would not have been available during an actual LNP. Also, there was no specific guidance provided regarding the need for flashlights to accomplish assigned tasks during the postulated LNP.
Plant personnel did not use personal safety equipment (hard-hats, gloves, etc.) during
evaluation of plant conditions following the postulated seismic event.
There were two examples of failure to document the TSC coordinator's approval for dose
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commitment on Form VYOPF-3507.02.
The licensee's expectations were not clearly defined regarding documentation of
dosimeter readings on VYOPF-3507.01 prior to leaving the OSC. The Radiation Protection service counter functioned as the OSC access control point for the Radiation
Control Area. Other OSC exit points did not have similar controls for dosimetry information.
There appeared to be some confusion between the requirements established in the
Emergency Plan Implementing Procedures (EPIPs) for the " Work Coordinator" and management expectations for this position. The licensee stated it intended this position to control in-plant activities, during an emergency, that fell outside the normal work control processes. Prior to VY's annual preparatory drill and this exercise, appropriate plant personnel received training that was consistent with established procedures.
However, during the drill and this exercise, the Licensee identified that the affected procedures needed further clarification. The licensee stated that it intended to establish longer term plans that would provide the needed clarification. This licensee-identified issue did not result in any adverse performance on the pan of the OSC organization during this exercise.
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2.6.6 Emergency Operations Facility (EOF)
No exercise strength or weaknesses were observed.
Overall, EOF management and control were good. Congestion and noise levels were kept low.
Logs were kept by the EOF principal staff. Radiological effluent and environmental monitoring and dose projections were performed. Status boards were generally kept up to date. Briefings were held on a reasonable frequency (about every 15 to 30 minutes). A second shift was
identified by 1050.
EOF managers and staff performed accident assessment and classification well. The Site Area Emergency was classified at 0647 based on automatic and manual scram signals and reactor
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power remaining greater than 2%. The Site Recovery Manager (SRM) delayed the classification of a Site Area Emergency by several minutes in order to allow time for the event to unfold to
make sure that the threshold for a General Emergency (GE) had not been reached. This delay was appropriate given the rapidly changing plant conditions. A GE was declared at 0850 based on the loss of two fission product barriers (RCS and containment) and the potential for the loss of the third (fuel clad).
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The SRM and his staff performed accident assessment continually and maintained a direct contact with the Plant Emergency Director (PED) and TSC. At about 0600, the SRM and his
staff aggressively pursued what turned out to be a simulator problem that resulted in a power increase when two control rods were inserted. From 0940 to 1025, the SRM and his staff continued to pursue the release pathways from the containment until they established that,
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indeed, the release was via the torus vent without any scrubbing effect. Tasks were appropriately prioritized and follow-ups made, and status reported.
The dose assessment staff, who arrived several hours before the release, used the time available to them to perform "what if" scenarios. Four field teams were used. The field team coordinator dispatched the field teams to be strategically located before the onset of a release. The dose assessment group used meteorological data and stack readings to make dose projections. The projections were compared with field measurements and showed good agreement. At least four air samples were taken by the field teams when one of these showed an I-131 reading of SE-8 uCi/cc at 0940, the dose assessment staff questioned the release pathway and whether the release was through the standby gas treatment system. Extensive discussions were held and drawings of the direct torus vent were studied. A backup system for dose projections, using a hand-held calculator (HP-41), was available and tested.
The field teams inquired at 0930 whether they should administer potassium iodide (KI). The i
EOF determined that, based on their dose projections and field measurements, KI was not needed.
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l Frequent communication between the different emergency response facilities was observed.
I Representatives from the three affected states (MA, NH, VT) were present. The licensee provided a technical team to brief the States and ensure their questions were addressed. The
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l SRM first briefed the States at 0555 (not all States had arrived). The EOF staff continued to
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verify that the TSC maintained a communication link with the NRC.
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EOF habitability was considered pnd proper instructions given to ensure its continued habitability.
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The licensee confirmed that off-site authorities understood the plant conditions and associated Protective Action Recommendations (PARS). The licensee also sought and received the status l
of protective actions implemented on a regular basis.
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Off-site authorities received an initial briefing on plant conditions and continued to be briefed
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throughout the exercise.
2.6.6.1 Core Damage Assessment i
Protective action was recommended to off-site agencies at 0852, within ten minutes after it was j
recognized that a GE threshold had been reached. Figure 1 of AP-3511 Revision 6 was used to determine the PAR. Using this figure, the SRM's staff concluded that there was substantial l
core damage in progress or projected. That determination resulted in a recommendation of t
evacuation of Vernon, Hinsdale and other towns five miles downwind. At this time, reactor
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water level was under control, high and low pressure emergency core cooling systems were l
available, and the core was covered. Containment monitors were reading about 170 R/hr, well l
below the level corresponding to substantial core damage. While a PAR is appropriate in a l
General Emergency, and the PAR made was appropriate in that it called for evacuation before a release began, the projection of substantial core damage was notjustified by the scenario data
(which indicated about 1% core damage). NRC review concluded that the Figure in AP-3511 l
Revision 6 lacked clarity to help the user to properly quantify " substantial core damage." This'
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was classified as an area for potential improvement. (IFI 50-271/93-07-03)
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2.6.6.2 Overall Event Response Timing l
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i The following table lists the times of Emergency Action Level (EAL) recognition, declaration and notifications, and the time of Protective Action Recommendation (PAR) issuance. The listed
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values show rapid review and promulgation of EAL-related data.
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EAL Recognition Declaration -
Notification PAR
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Unusual Event 0300 0305 0310
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Alert 0347 0350 0403
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Site Area Emergency OM5 0647 0653
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General Emergency 0848 0851 0900 0900,,
2.7 Media Center No exercise strengths, weaknesses, or areas for improvement were identified. it was observed that exercise questioners of licensee and State spokespersons were aggressive and posed challenging questions.
3.0 Licensee action on previously identified items Based upon discussions with the licensee, examination of procedures and records, and NRC observations, the status of open items is as follows:
The scenario aggressively challenged the licensee response in the Operations Support Center (OSC) where team briefings, team dispatch, and procedure adherence were much improved over the previous exercise. But, there were still some problems in this area. Therefore, OSC conformance to instructions and procedures, and the potential for inadequate plant configuration control in an emergency, will retInain open. (IFI 50-271/92-19-01)
4.0 Licensee critique On April 29, 1993 the NRC team attended the licensee's exercise critique. The Exercise Coordinator summarized the licensee's observations. No critique inadequacies were identified.
The critique was assessed as reasonably thorough and self-critical and identified most of the
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concerns identified by the NRC.
5.0 Exit Meeting On April 29,1993 The NRC team met with the liensee personnel listed in Detail 1 of this report. Team observatior.s were summarized.
The licensee was informed of the following:
Overall, the on-site licensee response to this exercise scenario was good.
- No violations were identified.
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The areas for improvement identified during this exercise were discussed.
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Licensee management acknowledged the findings and indicated that they would evaluate and take action, as appropriate, on the identified items.
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