IR 05000155/1987008
| ML20213A421 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/20/1987 |
| From: | Patterson J, Hironori Peterson, Snell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20213A418 | List: |
| References | |
| 50-155-87-08, 50-155-87-8, NUDOCS 8704280005 | |
| Download: ML20213A421 (16) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-155/87008(DRSS)
Docket No. 50-155 License No. DPR-6 Licensee:
Consumers Power Company 212 West Madison Avenue Jackson, MI 49201 Facility Name:
Big Rock Point Plant Inspection At:
Big Rock Point Site, Charlevoix, Michigan Inspection Conducted: April 6-8, 1987 Team Leader /,() $ M /.
- Mo/a7 Inspectors:
J. Pattersor Date h
H. Peters M
of//o/87 Date Approved By:
W. Snell, Chief 6d.C,.'A 4/ o/p7
Emergency Preparedness Section Date '
Inspection Summary Inspection on April 6-8,1987(ReportNo. 50-155/87008(DRSS))
Areas inspected:
Routine, unannounced inspection of the Big Rock Point Plant's emergency preparedness exercise involving observations by five NRC representatives of key functions and locations during the exercise. The inspection was conducted by three NRC inspectors and two consultants.
Results:
No violations, deficiencies or deviations were identified as a result of this inspection.
8704280005 870421 PDR ADOCK 05000155 O
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DETAILS 1.
Persons Contacted NRC Observers and Areas Observed J. Patterson, Control Room (CR), Operational Support Center (0SC) and Emergency Operations Facility (E0F)
S. Guthrie, CR, TSC, OSC H. Peterson, TSC and Post-Accident Sample System (FASS)
T. Lynch, TSC and EOF G. Weale, CR and TSC Consumers Power Company (CPC0) Personnel D. Hoffman, Plant Superintendent
- R. Abel, Production and Performance Superintendent R. May, Shift Supervisor
- P. Loomis, Emergency Planning Administrator, CPC0
- A. Katarsky, Exercise Coordinator, CPC0
- D. Fugere, Emergency Planner, CPC0
- M. Hobe, Emergency / Health Physics Superintendent Planning Coordinator
- J. Beer, Chemistry
- J. Horon, Shift Supervisor, Lead CR Controller
- G. Withrow, Engineering and Maintenance Superintendent
- S. Kiss, Property Protection, General Office, CPC0
- R. Krchmar, Quality Assurance
- D. McIntosh, Engineer
- C. MacInnis, Public Affairs Director
- D. Andersen, Controller-Palisades
- J. Brunet, Controller-Palisades M. Dawson, OSC Controller-Palisades G. Fox, Lead OSC Controller T. Elward, E0F Director, CPC0 R. DeWitt, E0F Officer, CPC0
- S. Amstutz, Secretary-Plant Superintendent
- Denotes licensee personnel who attended the exit interview on April 8, 1987.
2.
Licensee Action on Previously Identified Item (Closed) 0)en Item No. 155/86008-01.
This 1986 exercise weakness involved tie incorrect order for offsite notifications, in which the Alert was declared, contrary to 10 CFR 50.72(a)(ii)(3)gencies when the licensee notified the NRC before the State and local a During this
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exercise the licensee demonstrated the proper order for notifications for the Alert as well as the Site Area Emergency and General Emergency within the prescribed time limits. This item 1s closed.
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3.
General An unannounced off-hours exercise of the Big Rock Point Plant's Site Emergency Plan and Emergency Plan Implementing Procedures (EPIPs) was conducted on April 7, 1987.
The exercise tested the response of the licenseetoahypotheticalaccidentscenario,resultinginamajor
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release of radioactive material to the environment.
An attachment to thisreportdescribesthesco)eandobjectivesandanarrativesummary and chronology of events of tie scenario.
This was a partial exercise for the State of Michigan and a full-scale exercise for Charlevoix and Emmet Counties.
4.
General Observations a.
Procedures This exercise was conducted in accordance with 10 CFR Part 50, Appendix E requirements using the Big Rock Point Plant Site Emergency Plan and associated implementing procedures, b.
Coordination The licensee's response was generally coordinated, orderly and timely.
If these events had been real, the actions taken by the licensee would have been sufficient to permit State and local authorities to take appropriate actions to protect the public health and safety.
c.
Observers Licensee observers monitored and critiqued this exercise along with five NRC observers.
d.
Critique The licensee held a critique at the Big Rock Point Plant on April 8, 1987.
The NRC critique followed immediately after the licensee's self critique.
Personnel who attended this joint meeting are listed in Section 1.
5.
Specific Observations a.
Control Room (CR)
The CR operators demonstrated good responses to scenario messages, reacted cuickly and consistentl alarm anc system manipulations,y using proper procedures for all and anticipated future needs and
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impact of equipment losses.
Although there was good communication among the CR team, some messages and statements made by any given team member went unheard by others whose attention was concentrated on a specific emergency issue.
Some oral acknowledgement or request to repeat the information would be helpful in this crowded, occasionally noisy environment.
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The'CR team aggressively. pursued alternative actions and. options throughout'the exercise.
Examples included the rerouting of~
offsite power'after it was completely-lost about 0800 and the finding of alternate sources of water supply for the Condensate.
Storage Tank.
Good log keeping was maintained in'the Control.
Room.
From a CR viewpoint, more' systematic-~ involvement of Health Physics (HP) representatives-in monitoring dose rates and addressing.
.ALARA considerations for the CR was warranted.
The Shift Supervisor (SS) spent approximately ITof the first-21. minutes after the-Alert declaration in preparing and sending emergency notifications.~ During this time he was not in position to supervise the plant response-to the LOCA or the general site response-to the emergency.
For the. Alert notification he did not provide the Counties =with much of the data listed on-Attachment 2 to EPIP-6F, Emergency Notifications.
Notifications were made to' the.
State, Counties and the NRC within the, time requirements and;the-proper sequence for both the. Alert and the Site Area Emergency (SAE).
Completion of the SAE.offsite~ notifications and updates were
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made by the TSC Communicator; thus allowing the SS to concentrate.
more on_his CR responsibilities.
The Plant Announcements-(PA) of-the Alert and the SAE did not include any'cause or brief descriptive statements of the reason for declaring the emergency.
This should always be included to inform those in the plant or personnel.in other buildings or outside the reactor building who are not part of-the emergency response organization.
The On-Call Technical Advisor (OTA) primarily functioned as a data transfer agent between the CR and TSC.
He did not appear to be accomplishing his responsibilities as defined in Procedure EPIP-4C.
Some of these responsibilities he could have pursued included calculating core damage and release rates, and verifying completion of mandatory actions for emergency classifications.
Also, he should have been involved in identification of conditions and alternative solutions to mitigate the accident.
b.
Technical Support Center (TSC)
The TSC was effectively staffed, with the exception of'the Data.
Recorder, within 48 minutes after the two-minute siren and the-accompanying Alert level declaration. The TSC effectively assumed their operations at the advent of the declaration of the SAE. -Prior to this there was no official announcement made when the Site Emergency Director (SED) took over command and control from the Control Room.
There was discussio'n_and apparent agreement ~
between the two directors.
However, no documentation was found in the TSC log book to verify the-transfer of authority.for control-of the emergency to the TSC.
Supporting staff; assumed their-responsibilities in an efficient manner and expeditiously began to implement their duties.
Notifications and coordination with the State, Counties, NRC, and such other organizations as offsite fire departments and contractors were well demonstrated.
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Assembly and accountability, initiated at the Alert level, was controlled by the Site Accountability Coordinator at the Security Building.
The report received at the TSC confirmed that this was completed within 22 minutes after the Alert siren was sounded.
The SED arrived 45 minutes after the Alert was declared, somewhat late for the established 30 minute response time goal for that position.
The SED provided good briefings, discussions, and coordination with his TSC support team leaders.
Among important areas discussed were possible corrective measures, potential plant trouble areas, precautionary actions and Protective Action Recommendations (PARS).
The volume of the plant PA system within the TSC was often too low and not heard by everyone in the TSC.
The SED could have repeated the information and addressed it directly to the TSC staff.
Plant status boards, plant parameter trend graphs and the chronology of significant events were slow in being utilized.
It was almost'
40 minutes after the SED arrived before information was entered on
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the status boards.
The events board was not posted until approximately one hour and 20 minutes after the SED arrived.
One status board was not updated from the time power was restored at 1000 to indicate restoration of Core and Containment Spray until approximately 1130.
For other times in the exercise status boards were adequately maintained.
When the HP technician opened the unsealed emergency kit containing radiation monitoring equipment, no effort was made to take radiation source checks or operational checks on the instruments before using them. This should be done routinely whether for inplant or offsite use.
The HP technician next established radiation boundaries and control points with frisking equipment to control access to and from the TSC.
He periodically took air samples and does rate surveys.
Access control was not strictly followed by those leaving or entering the TSC approved passage.
Several participants failed to frisk prior to TSC entry.
The documentation form, provided to keep records of exposure control for personnel entering the plant from the TSC, was not used. An informal method using a sheet of paper listing the variables and times involved was used, but not consistently.
Smears were taken without documenting where they were taken.
Filters were placed on top of the frisker and handled without using gloves.
Initially, the dose assessment actions and offsite monitoring activities were handled in an attentive and timely fashion.
However, the EPZ map was not updated as effectively as it should have been.
For example, the affected sectors, survey locations, and tne affected sectors for evacuation were not posted on the map.
Only arrows indicating wind speed and direction were put on the map.
No changes were made until the Security Officer and maintenance crew called to ask for directions and radiation levels for evacuating personnel.
Team location and dose rates were listed on the radiological status boards; but the group responsible for offsite plume tracking and other dose assessment related responsibilities should be trained to utilize the EPZ map correctly and continuously.
Even after the E0F took over this responsibility, the EPZ map in the-TSC should be kept current to better provide backup and analyze conditions in concert with the E0F.
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Communications were well maintained withithe CR, OSC and E0F'throughout this 10-11 hour exercise with minor. exceptions. The initial offsite'
-power loss'at 0800 resulted in.lo'ss'of some telephone and facsimile.
capability.
Some commercial telephone lines were not affected by.the-
. power. loss..Overal1(theTSCreactedquickly'toovercomethispower loss.
Hand-held radio communications proved to be.a problem especially'.
as more groups including the offsite monitoring teams, security, Land:
others realized that-they had the same frequency for transmission.
-This created some confusion and delays. More channels ^must be<used,
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in the future to keep up with the demand for emergency response use.
The flow of pertinent data between the CR and TSC was timely and well '
maintained.
The TSC logs were.kept current including'the SED'stlog:
which was continually updated by the Data Recorder.
The SED routinely-updated his counterpart,_the E0F Director.
There was a smooth;
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- transition of command and control to the E0F after the EOF Director determined that he was adequately briefed on plant conditions by-the SED.
Based on the above findings, the following items should be considered for improvement:
The Site Emergency Director should announce and document the
activation of the TSC when he' assumes command and control from the Control Room.
The Health Physics Group should more effectively utilize and
annotate the EPZ map especially before the E0F takes over this function.
Appropriate forms should be used by HP to document personne1'
exposure records.
Two-way hand-held radios should not be limited to one frequency
for transmission. Too many groups using the-one frequency resulted in poor reception and confusion to the emergency response personnel, deterring effective communication.
c.
Operational Support Center (OSC)
The OSC was declared activated and functional at within approximately one hour after the Alert was declared.
Reactor and containment conditions were not posted as soon-as they should have been after OSC initial activation.
Early in the exercise these. status board entries lagged behind real time for up to 45 minutes.
This improved i
with the addition of a data recorder, which reduced the lag to-ten minutes.
Plant drawings, references and other eguipment related information were available and well utilized.
Briefings b OSC Director were substantive and frequent using_a " bull" y the horn to overcome.the noise background.
Dispatch of several repair teams to the Screenhouse to examine, evaluate, and provide.a new filter were conducted with a good interchange of possible solutions and involvement of maintenance personnel, C&RP technicians, plus an auxiliary operator.
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The demonstration of actually obtaining a replacement. filter for the diesel driven fire pump from the stockroom and bringing it to the Screenhouse produced numerous deficiencies.
Problems with computer access, keys and then the selection of the wrong sized filter-resulted in time consuming delays.
Originally this event sequence was scheduled.to be simulated, but was demonstrated at the request of the NRC observer, thereby showing the need to carry out actions rather than simulate them.
Personnel accountability and exposure level--tracking of OSC teams and others was well maintained throughout the exercise.
Communications with the TSC were well maintained.
Radio communications were erratic at times. Whatever transmission frequency the Chemistry and Radiation Protection Supervisor used from the OSC; it was able to overcome and dilute the reception between other two-way radios.
Reception from Team 1 of the offsite teams was also badly impaired.
(See Section 5.e).
The OSC location in the air compressor room directly adjacent to the machine shop is a very difficult environment in which to function effectively.
Loud continuous background noise, heat, and lack of ventilation did not help the OSC staff to function.
One Controller suggested the use of portable panels of insulating, sound proofing material that could be temporarily fastened around the compressors and related equipment.
Some adjustment would be beneficial.
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Post Accident Sampling System (PASS)
The scope of the exercise scenario did not incorporate the necessity.
of a PASS sample.
According to the EPIPs, the PASS procedure is utilized for estimated core damage of between zero to ten ?ercent.
The exercise scenario was set for core damage of greater t1an-20 percent.
Therefore, the PASS sample was demonstrated after the termination of the original exercise utilizing a separate scenario developed for the PASS.
The sampling team consisted of three individuals, two to take the sample, and the third to act as the supervisor.
Prior to sample collection activities, the team members thoroughly reviewed and
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discussed the relevant procedures.
They also demonstrated good ALARA precautions and practices. They planned out their actions, task assignments, routes, and estimated their dose to obtain the sample.
There were some inaccuracies in recording the dose units on the procedural form. The form was written for dose units in Rem, but the dose recorded was for mrem.
Thus, the dose of 400 mrem was mistakenly recorded as 400 Rem.
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The teams ability to select proper equipment, clothing, dosimetry and to don the anti-contamination clothing was adequate.
They observedgoodHPpracticesuptojustpriortodrawingthesample.
Once the simulated sample was collected, the demonstration of
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removing anti-contamination clothing and proper H P. precaution in handling the. reactor coolant sample was insufficient. 'The use of gloves was not demonstrated in transferring and diluting the sample.
Due to the plant configuration there was no actual reactor coolant-sample drawn,:it was simulated. ~The PASS is:a.3/4. inch drain line-on the core spray heat exchanger.' This drain line penetrates.
outside the containment sphere.
The drain:line valve is chained and locked, and is considered part of primary containment.
Thus,'during-
. plant ~ operations, the drawing of a PASS sample is simulated.to
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preclude from breaching containment integrity.
Although the actual sampling may be simulated, there was an excess.
of simulation.
Examples of simulation included the dilution and:
transferring of the sample; the use of shielding for..the storage of the sample; and lack of water in the containers to: represent the reactor coolant. The actual counting of the sample and operation of the computer was also' simulated. -All of these actions could have-been practically demonstrated.
Overall, the majority of the actions were simulated, but the technical knowledge and PASS capability was, in part, adequately demonstrated.
Based on the above findings, the following' items should be considered for improvement:
A better scenario which emphasizes more' actual. demonstration
and less simulation.
As conditions allow, demonstrate actual drawing of a reactor coolant sample using the PASS.
- Better attention to details in-the demonstration of radiation protection techniques including donning of anti-contamination clothing and data recording.
e.
Emergency Operations Facility (E0F).
The loss of offsite power was announced at the E0F at 0800.
The staff responded by determining ways to compensate for the loss, both for communications and loss of all AC power since diesel generators were unavailable.
The E0F switched to-dedicated, independently powered telephones to maintain vital communicators during this power outage.
The HP Support Team recognized that. fuel failure indications could be identified using portable instrumentation located at the Security Building.
It was subsequently determined that a portable instrument could be located near containment with a readout in a
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shielded location which could also be used to indicate fuel failure using EPIP-50. Throughout the E0Fs activation the support: teams were very responsive and reacted quickly.and accurately, in most cases to
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anticipatepossiblebreakdownsandnotjustsolutionsafterthe damage was.done.
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'The E0F Director, E0F Officer, EOF Administrator HP-Su
~ Leader.andtheEmergencyPlanneroftheGeneraldffice'pportTeam.
Response Team:
L (G0RT) arrived at 0620 at the Boyne-City, Michigan E0F building.. This
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-was approximately two hours and 23 minutes after the.SAE was declared.
The Big Rock Point Site Emergency Plan,1Section 9.7, states.that the
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goal for the G0RT team to arrive is about 11/2 hours.
Contact was-
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being maintained, however, by telephone with the TSC and E0F while the plane was in flight to Charlevoix.
Thus the.E0F Director and his management staff were apprised of. emergency conditions on a
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.real time basis while.enroute.
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f Briefings by the E0F Director'were frequent and meaningful. He frequently met with the.various support' team leaders and held caucuses to discuss proposed actions and get' input from the teams.
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Four dedicated Communicators provided continuous csntact with the State, Counties, outside suppliers and other support agencies. This
communication group'did a good job in maintaining contacts and recording messages.
The offsite-field teams' data were utilized to calculate a release rate during the The source term value calculated from the 15 mR/ power outage.
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hour reading taken at the1 site boundary
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i indicated that the leak rate exceeded design basis leakage.
Forecast
meteorological data was considered by the E0F Director to anticipate which sectors of the EPZ map would be affected if there was a' shift
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in wind direction.
Dose assessment values were obtained on a timely
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i-basis using the IMB PC computer with printout.
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The dispatching of the two field teams were observed from a f
ccmmunications and logistical standpoint, since no observers were-l assigned with the teams. Only a single traverse of the plume was
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made beyond three miles as determined from communication with the
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teams from the E0F.
To characterize the extent of the plume more i
information than this would be necessary.
Teams No. 1 had radio
transmission problems about 0930.
Team No. 2 was used to relay j
information to the E0F. The HP Supervisor initiated instructions to Team No. 1 on how to determine if readings were due to " shine"
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from containment or from the plume.
PARS were correctly issued by
the E0F Director after consulting with his HP Support Group Leader.
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Communications with the State HP counterpart appeared to go well.
A good dialogue between the licensee and State was apparent from the E0Fs vantage point as observed _by the inspector.
Reentry / Recovery (R/R)
plans could have been more comarehensive and meaningful than.was demonstrated.
At about 1215, UR was being considered as directed by i
the Emergency Director and Emerger, icy Planner.
Each support' group was i
asked to contribute, based on their area of expertise and following j
.the guidelines of EPIP-6.C, Reentry / Recovery.
There was a brief
meeting of the group leaders.in a separate office to discuss the L
subject. 'At 1223 the emergency was deescalated to an Alert. The l
Plant Support Engineering Group Leader did not appear to understand
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what was expected of his group after receiving a printed outline e in EPIP-6.C.
Also the TSC did not appear to be ready from a pag /R step when contacted by the E0F Director. ~This final i
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phasewasaddressed,andinoutlineformmettheexercise. objective, but more planning and scenario involvement is needed to make R/R morerealisticandobjective.
6.
Exit Inte'rview The inspection team held an exit interview on April 8,1987 with those
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licensee personnel identified in' Paragraph 1.'
The Team Leader discussed the scope and preliminary findings of the inspection.
The inspectors determined from the licensee that none of the information discussed was considered proprietary.
Attachments:
1.
ExerciseScopeandObjective.
2.
Exercise Narrative Summary
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1.0 SCOPE AND OBJECTIVES 1.1 SCOPE BREX 87 is designed to meet exercise. requirements specified-in 10 CFR 50, Appendix E, Section IV.F.
It will postulate events which would require activation of major portions of the site emergency planLand response by offsite authorities. The exercise will be unannounced to Consumers Power Company,' State of Michigan and County emergency response personnel.
It
.will also occur during the_ area's winter season. Real time activation of Company and offsite emergency response facilities will occur.
1.2 OBJECTIVES The exercise will demonstrate:
1.
Assessment and Classification a.
Recognition of emergency conditions b.
Timely classification of emergency conditions in accordance with emergency action levels
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2.
Communication a.
Initial notification within specified time constraints (State and local - 15 minutes, NRC - I hour)
b.
Subsequent notification in accordance with procedure c.
Notification and coordination with other organizations, as required (other utilities, contractors, fire or medical services)
d.
Provision of accurate and timely information to support news release activity e.
Provision of rumor control support 3.
Radiological Assessment and Control a.
Calculation of dose projection based on sample results or monitor readings b.
Performance of in-plant and offsite field surveys c.
Collection and analysis of a post-accident primary coolant sample d.
Trending of radiological data e.
Formulation of appropriate protective action recommendations mil 086-0006A-TPil-TP15 1.1 L
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Contamination and exposure control 4. ' Emergency Response Facilities
a.
-Activation, staffing and operation at appropriate classifications and within specified time constraints b.
Adequacy of emergency equipment and supplies c.
Adequacy of emergency communication systems i
d.
Access contro.1 j
5.
Emergency Management
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Command and control with transfer of responsibilities from Control Room to Technical Support Center to Emergency Operations i
l Facility b.
Assembly and accountability within 30 minutes
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c.
Coordination with State of Michigan emergency organization d.
Mitigation of operational and radiological conditions
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Mobilization of emergency teams e.
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6.
Reentry and Recovery t
a.
Assessment of damage and formulation of recovery plan
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identification of constraints, requirements and organization to implement the plan j
7.
Exercise Control
t a.
Provision for maximum free play
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Accurate assessment of player performance
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mil 086-0006A-TP11-TP15 1.2
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3.0 SEQUENCE OF EVENTS 0300 - APRIL 7, 1987 The reactor has been on-line for several months operating at= 90% rated power.
A diesel fuel delivery made within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to top off all diesel tanks has resulted in contaminated fuel in each of the diesel systems (fire pump, emergency generator and standby generator). Operations is unaware of this situation.
Primary coolant has begun to slowly leak to the reactor cooling water system (component cooling) by way of the clean-up system n'onregenerative heat ex-changer through a small, but slowly increasing, tube leak.
At 0315 the leaking clean-up heat exchanger tubes abruptly rupture causing a small Loss of Coolant Accident (LOCA) condition to occur. Co'ntainment dew cell and temperature sensors rise off scale. Containment pressure rises. The reactor scrams and containment isolates at 1 psig enclosure pressure.
An Alert should be declared based on indications of a LOCA. The Technical Support Center and Operations Support Center are activated. Plant personnel are dispatched to the Emergency Operations Facility (EOF). The General Office Response Team (GORT) in Jackson is mobilized.
The EOF will be fully activated based on the judgement of the Site Emergency Director and General Office Management. Upon confirmation of plant conditions, the GORT will likely, opt for full EOF activation.
The reactor depressurizes relatively rapidly and enclosure spray actuation oc-By 0330, the containment pressure peaks and reactor depressurization is curs.
nearly complete. The diesel fire pump fails to start but the electric pump is available if the operator chooses to manually actuate the core spray.
At 0800 a loss of offsite power occurs. Reactor level drops and core spray fails as the diesel pump is under repair and the EDC fails to start. A Gen-eral Emergency is declared based on these conditions once reactor level drops to the top of the core. Various operator attempts to reestablish core spray flow by way of the standby diesel generator and the diesel fire pump are frus-trated as none of the diesel equipment will not run very long due to the contaminated fuel supply.
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Upon declaration of a General Emergency, protective action recommendations are made to offsite authorities.
At 0930 radiation levels are noted to rise on hand-held instrumentation due to gradual cladding failure. The core is uncovered for+ 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> as attempts to repair the diesel equipment fail. At 1030 an offsite power source is estab-lished and the electric fire pump is returned to service. However, the return of ac power causes a slight power surge on instrumentation buses. This surge fails vacuum relief instruments in the containment exhaust vent path resulting l
in a false vacuum relief signal opening the vacuum relief valves. Containment i
atmosphere, containing fission products released from the uncovered core, is now being ventilated to the stack.
PR0187-0313A-TP15-TP02
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I&C personnel disable the faulty vacuum instrumentation, but debris blown around the containment during the LOCA is lodged in the vent pipe preventing valve closure. I&C continues to exercise the valves with false vacuum relief signals until one of the valves is managed to be closed at 1145. The release terminates at this point.
Plant conditions stabilize. The plume clears the EPZ at about 1200. The emergency deescalates. Reentry and recovery efforts commence. The exercise terminates at 1415.
PR0187-0313A-TP15-TP02
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'k Narrative Summary
Scenario Time Key Events 0015 Dew cell alarms occur. Reactor pressure takes a sharp turn downward.
0017 The reactor scrams on high containment pressure.
0022 Reactor pressure is ~450 psi and falling. Containment pres-sure is rising. Feedwater at full flow trips on low suc-tion.
(Hot well is depleted.)
0025 Diesel fire pump fails to run due to fuel contamination problems. An alert'is declared due to indications of a LOCA. Follow-up notifications are made to offsite agencies and authorities.
0030 Containment pressure rises to 10 psig and levels off. Reac-tor pressure is ~250 psig with reactor water level gauge full scale high.
0045 A feed pump is returned to service, but cannot regain the steam drum level before tripping on low suction. Reactor pressure crosses through 100 psig (and core spray can be manually placed in service if the operator chooses to do so).
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0100 Reactor pressure 75 psig and falling. Containment pressure 8 psig and falling. Drum level full scale low; reactor level full scale high.
If core spray was not placed into service at 0045, feed pumps can be restarted at this point to maintain level above the low reactor water level set point. 3,000 gallons (normal hot well inventory) will be replenished by gravitational feed from the condensate stor-age tank in a 15-minute period, which is sufficient to per-mit a feed pump to run at rated flow for 2 to 3 minutes.
0145 Reactor pressure 50 psig and falling. Containment pressure 7 psig and falling. Reactor level full scale high; drum level full scale low.
If feedwater has been used to main-tain primary system inventory, then the condensate storage tank is empty by now. In order to maintain reactor level, core spray must be placed in service.
0245 Reactor pressure 30 psig and falling. Containment 6 psig and falling. Reactor full scale high; drum full scale low.
0330 Reactor pressure 25 psig and falling. Containment 5 psig and falling. Reactor full scale high; drum full scale low.
0415 Reactor pressure 20 psig and falling. Containment 5 psig and stable. Reactor level full scale high; drum full scale low.
PR0187-0313A-TP15,-TP02 L
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'l Narrative Summary
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Scenario Time Key Events 0500 Reactor pressure 18 psig and falling. Containment 4 psig and stable. Reactor level full scale high; drum level full scale low. A loss of offsite power occurs. Core spray flow drops to O gpm (as diesel fire pump is still under repair and EDG will not run due to fuel problems).
0505 A0s dispatched to station power room and standby diesel to
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transfer emergency bus to standby diesel. Standby diesel
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fails within a few minutes due to fuel problems.
0515 Reactor level near full scale high. Reactor pressure rising to 50 psig (reading available at alternate shutdown panel).
Containment pressure rises to 5 psig.
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0530 Reactor level in lower half of reactor level instrument scale.
0540 General emergency declared. Core uncovery is imminent.
0545 Various attempts by operators to regain core spray.
0600 Containment and reactor pressure continue to rise gradually.
0630 Radiation levels begin to rise via hand-held instrumenta-tion.
(Fuel gap release is beginning to occur.)
0700 Area radiation levels continue to rise. Localized fuel melt begins.
0730 Power to the 46-kV line is restored. A power surge causes circuitry in the vacuum relief system to fail causing a false vacuum relief signal and failing a train of vent valves open. Release occurs
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0845 Relief valve closed by I&C.
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0900 Plume clears EPZ. Reentry and recovery begin.
1115 Exercise ends.
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