ML19305D542
| ML19305D542 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 04/09/1980 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 LAC-6853, NUDOCS 8004150204 | |
| Download: ML19305D542 (15) | |
Text
e g-D DA/RYLAND COOPERAT/VE PO box 817 2615 EAST AV SOUTH
- LA CROSSE. WISCONSIN 5160t a.608) 7284000 April 9, 1980 In reply, please refer to LAC-6853 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation ATTN:
Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D. C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 ADDITIONAL INFORMATION - THREE MILE ISLAND SHORT TERM LESSONS LEARNED (NUREG-0578)
Reference:
(1)
DPC Letter, Linder to Denton, LAC-6769, dated January 31, 1980.
Gentlemen:
The attached information is in response to requests for additional information resulting from an NRC site review of compliance with NUREG-0578.
Included are the revised pages to Reference 1 which were required as part of this additional information.
If you have any questions regarding this submittal, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE 14 i
Frank Linder, General Manager FL:JDP:af Attachment cc:
J. Keppler, Reg. Dir., NRC-DRO III 800.4150 p
4 NRC REQUEST:
2.1.3.b(1)
Provide a deceription of the ongoing program of neu instrumentation to detect inadequate core cooling conditiona.
Modify page 4 of January 31 letter to reflect thia action is not complete.
DPC RESPONSE:
DPC believes that LACBWR now has sufficient instrumentation for detection of inadequate core cooling as discussed in DPC letter LAC-6769 dated January 31, 1980.
DPC's I&E Department has an ongoing program to keep abreast of new developments and state-of-the-art techniques which may have potential application as improved water level instrumentation for BWR's.
The I&E Department will evaluate any new instrumentation which offers significant advantages over existing LACBWR equipment, and will present its findings to the ORC and SRC for consideration.
NRC REQUEST:
2.1. 4 (1)
Modify procedurea that require the operator to set the valves in closed position prior to react of the isolation signal.
DPC RESPONSE:
The modifications specified were reviewed, as part of IE Bulletin No. 79-08 review, and in cases where the procedures did not call for such action, it was included.
(2)
Commit to perform a revicu of the overall design of isolation function to determine any modification that could eliminate th* above procedural control.
The above isolation function of the containment building damper circuit has been reviewed.
It has been concluded that the circuit which will be modified to prevent automatic opening of the dampers whenever the closure circuit has been cleared is adequate.
Further modification would overly complicate the circuit, new switches and extra relays would be necessary and this could be the source of a malfunction of the equipment.
(3)
Provide a list of valuco on a ayatem bacio that has the capability to override an iaolation aignal.
There are none.
Valves in air lines would only isolate the control
4 i
NRC REQUEST 2.1.4 - (Cont'd) air which would insure that the valves fail in their " isolated" position.
All safety related valves fail in the closed or safe direction on:
1.
Loss of air 2.
Loss of power 3.
Isolation signal.
(4)
Provide a list of systema which you might need for potential recovery from an accident.
Describe uhat methoda vill be used to reopen these systems if the isolation signal sustains.
The non-essential systems (identified on Pg. S&6 of LAC-6769, DPC Letter, Linder to Denton, dated January 31, 1980), which receive diverse parameter isolation during an event which might be utilized in post-accident recovery situations are:
1.
Decay Heat Removal Startup Line.
The use of this system to remove excess primary system coolant for processing or storage outside the reactor contaimnent building would not normally be permitted if either a low reactor water level signal or a high containment building pressure signal were present.
However, in the event of building flooding to core midplane, a high building pressure signal will be present.
Utilization of the decay heat blowdown line to reduce containment water inventory would require jumper installation around the high building pressure isolation signal.
If the signals are due to a single instrument mal-function, the low reactor water level signal could be bypassed by use of a key controlled bypass switch, and the high containment building pressure signal would require the use of jumper cables for bypassing.
2.
Containment Building Drain Line.
This system is isolated by two locked closed manual valves outside of containment.
No bypass actions are required.
NRC REQUEST:
2.1. 5. a (1)
Revise your January 31 response on pagca 8, 9,
and 10 that Hg purge ayatem ic not required as part of the licenced baaca of your plant.
DPC RESPONSE:
A revision of pages 8, 9, and 10 of the January 31, 1980 response (DPC Letter LAC-6769), identifying that the purge is not a license bases of our plant, is being submitted on the following three pages.
Reviecd Pagn E
Mr. Harold D:nton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems i
and 2.1.8.a Improved Post-Accident Sampling Capability Pont-accident purging has not been included as a design basis for licensing LACBWR under Provisional Operating License No. DPR-45.
No modifications are required.
A review of existing post-accident sampling capabilities at LACBWR against the requirements of Section 2.1.8.a of t
NUREG 0578 has been completed.
Existing sampling provisions do not meet the requirements, since currently available sample locations would be inaccessible in the post-accident scenario postulated by the Task Force.
s The facility modifications required to bring the plant into full compliance have been identified.
A new, additional remote sample station will be installed at an accessible location outside of the containment building.
Procedures are available to perfomm boron and chloride chemical analyses assuming a highly radioactive initial sample.
Full implementation of the post-accident sampling system will be ccmplete by January 1, 1981.
With respect to improved post-accident sampling capability, liquid reactor coolant samples will be conveyed from the sampling point in containment to the designated post-accident sampling station in the turbine building via 3/8" stainless steel tubing.
The sample line will be equipped with:
a manual isolation valve close to the source, a remote-manual containment boundary valve, a sample cooler, a remote-manual flow control pressure reducing valve, pressure breakdown orifice, sample cylinder, miscellaneous check valves and flow directing valves, and instrumentation to measure sample flowrate, pressure and temperature.
Sample fluid will flow in a closed loop from the sample point, out to the post-accident sample station, and back into containment where it will be directed to the reten-tion tanks.
Provisions will be made to introduce demin-eralized water into the sample stream to dilute it, if
.. ~. -
pg.Lecd Page Mr. H rold Danton, Director LAC-6769 Of fico of Nuclocr Rzactor Regulation January 31, 1980 t
Section f
I No.
Title 2.1.8.a required for safe handling, before it reaches the sample (Cont'd) station.
Samples will be obtained by:
a) purging to the retention tanks until fluid representative of the recirculating reactor coolant fills the line; b) adding demineralized dilution water to the sample stream to the degree required; c) after equilibrium is reached, divert-ing the sample stream to a sample cylinder, which is sub-sequently isolated, manually removed, and transpo*ted to the health physics laboratory for analysis.
Post-accident hydrogen content of contaimment air will be monitored by continuously passing a small sample stream of the containment atmosphere through a hydrogen analyzer located outside of containment.
A small compressor will provide the motive force to draw sample air from contain-ment, route it through the analyzer, and return it to containment in a closed loop.
Continuous readout of H2 concentration will be provided in the control room.
A l
block valve will be installed downstream of the sample A bypass line around this valve will contain compressor.
a sample cylinder.
A discrete containment air sample can be obtained manually by bypassing the circulating contain-ment sample air through this line, isolating the sample cylinder, and removing it to the laboratory for analysis.
Manual sampling will be performed in a shielded, accessible area in the turbine building.
Full implemer.tation of sampling capability and hydrogen monitoring will be completed by January 1, 1981.
2.1.6.a Integrity of Systems Outside Containt.ent Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs.
At LACBWR, all systems which would contain large inventories of radioactive materials following a serious accident or transient are located entirely inside containment.
As a result, the special leak measurement and leak reduction measures described in Section 2.1.6.a of NUREG 0578 are not applicable to this plant.
DPC will continue to strive to maintain leakage from components in normally radioactive systems to as low as practical levels.
Leakage detection is normally achieved with the plant on line by monitoring local and stack i I
Ravised Pagn Mr. Hcrold Danton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.6.a activity which is indicated and recorded continuously in (Cont'd) the control room.
Since all plant building and equipment off-gas vents are routed ultimately to the stack, there are no locations where leakage could continue undetected.
At LACBWR, maintenance of systems with indicated leakage has always been accomplished on an expedited basis.
L 2.1.6.b Desic n Review of Plant Shielding of Spaces for Post-Accic;ent Operations l
In accordance with the position outlined by NUREG 0578, l
and clarified by Reference (2), a radiation and shielding design reivew of areas which will require occupancy by an operator to aid in the mitigation of or recovery from an accident has been performed.
Three areas have been identi-fled in which access will be necessary:
The control room, I
the post-accident sampling station, and the health physics laboratory.
The design review has found that no new sources, except the samples obtained, would be generated in any of these areas as the result of a post-accidental release of radioactivity equivalent to that described in t
i
NRC REQUEST:
2.1. 6. a (3)
Address your work to respond to our letter concerning the North Anna and related incidents probicm.
DPC RESPONSE:
A review of waste transfer procedures was conducted to determine if significant attention to transfer was prescribed.
The pro-cedures are adequate.
The design of LACBWR waste tanks and reten-tion tanks (which are inside containment) provides a high level alarm in the control room if the operator were to fail to terminate the transfer.
If the alarm was overlooked, the tanks overflow into each other (rather than the building floor).
The vents on both retention tanks and both liquid waste tanks are connected to a vent header which is monitored prior to release.
No apparent potential for any event identified in IE Circular No.
79-21 was found.
No modifications are proposed.
NRC REQUEST:
2.1. 6. b (2 )
Discuss the radiation qualification of the ESF equipment in the plant.
Refer to any documentation that has been previously submitted.
If not adequately addressed, provide that information nou.
Due prior to June 1, 1980.
DPC RESPONSE:
Please refer to the following documents:
1)
Environmental Qualifications of Safety Related Equipment:
DPC Letter, LAC-5181, to NRC, Madgett to Stello, dated February 22, 1978.
NRC Letter to DPC, Ziemann to Madgett, dated September 6, 1978.
DPC Letter, LAC-5520, to NRC, Madgett to Ziemann, dated October 26, 1978.
9 NRC REQUEST:
2.1. 8. a (1)
Document that you have procedures to take and analyze camples (RCS and containment air) uhich take into account high radiation camples.
These samples vould be taken if the containment is accessible.
DPC RESPONSE:
Emergency Plan Procedure #21 (EPP-21) describes the considerations for obtaining high activity liquid and gaseous samples during emergencies, including during access to the containment building.
Various Health and Safety Procedures (HSP) describe the methods to obtain these samplesand the considerations necessary when analyzing high activity samples.
NRC REQUEST:
2.1.8.a(2)
Document where the existing RCS sample tapa are located on the RCS and that they vill provide a representative sample during an accident.
State if these vill be the same ones (and only ones) to be used for the neu cample station.
DPC RESPONSE:
The existing sample taps on the Reactor Coolant System are located as follows:
a)
Primary Purification Cation Bed Inlet b)
Primary Purification Cation Bed Outlet c)
Primary Purification Mixed Bed Outlet d)
Decay Heat Pump Vent e)
Various Reactor Vessel Level Probes f)
Additional vents, drains, pressure transmitter fittings on Main Steam Line, Feedwater Line, Shutdown Condenser System.
Location "a,
b, and c" all tap directly off the bottom of the reactor vessel and would be the normal sample points.
Location "a" would be most representative if the Purification System was not isolated.
If it was, location "d" would be representative of the Forced Circulation Pump loop on the reactor side of the isolation valve.- Location "e" would be used if "a through d" were not accessible.
Locations "f" would be used if the Reactor Building was not accessible.
The new remote sample station will tap off an isolable point on the vessel as presently conceived.
NRC REQUEST:
2.1. 8. a (3)
Describe off-gaa high radiation monitor, and its location near the containment, uhich could be used to monitor radiation from containment and, through procedures, the grosa concentration of activity in containment.
Describe the means to quantify high activity releasca through the off-gas syaten.
DPC RESPONSE:
The Victorcen Fuel Integrity Monitor is located in the Electrical Penetration Room near the biological shield and mounted on the off-gas line at the effluent of the 10-minute holdup tank directly across from the pipe Tracerlab Off-Gas Monitor.
The shield at this point between the reactor building and penetration room is 9 inches of concrete and 11.16 inches of steel.
Assuming a 1200 R/hr dose in the Reactor Building, the dose rate to this monitor would be 30 Rem /
7 hr from direct radiation.
Since the monitor reads to 10 R/hr, it could be used for off-gas monitoring during emergencies.
In addition, as stated in response to audit item 2.1.8.b(2), additional lead shielding will be placed in this area.
Contrary to the idea ex-pressed during the audit by the LACBWR staff, it does not appear feasible to use this monitor for the Reactor Building radiation field estimates at design LOCA dose rates.
However, #12 area monitor which is located on the 24-inch thick south Control Room way common to the Reactor Building would be usable for estimating the Reactor activity.
Corrections would be made based on the FSAR radiation values during a LOCA for sky shine and direct doses.
Various procedures discuss this monitor.
NRC REQUEST:
2.1. 8. b (1)
Describe the monitor that vill be used to quantify high level grosa activity being released via the atack during an accident.
Provide commitment to install this monitor uithin tuo oceks of delivery but no later than prior to startup from the April, 1980 outage.
DPC RESPONSE:
The monitor in question is called the Hi-Range Stack Noble Gas Monitor.
It is an Eberline Model HP-270G gamma G-M detector installed in a lead shielded pig monitoring the Tracerlab Stack Gas Monitor dis-charge piping.
It was installed on March 31, 1980.
It will be collimated and calibrated upon startup when the present outage is completed although some data has already been obtained.
It is connected to an Eberline Model RM-19 chassis located on Panel G in the Control Room.
Its use is described in EPP-21.
NRC REQUEST:
2.1.8.b(2)
Provide data requested in the October 30th letter, pagcc 33, 34 and 35.
DPC RESPONSE Range KEV Calibration Name Iocation (uC1/cc)
Eneruy Range Frequency Technique Tracerlab Off-Gas, lO-Minute 10-4 to 60 - 2000 Monthly Grab Sample Model PO-5B Holdup 10+2 Analyzed on NaI Crystal Tank GeLi Detector Effluent Victoreen Fuel 10-Minute 10-3 to 70 - 3300 mnthly Grab Sample Integrity, Holdup 104 Analyzcxl on 2 del 847-1 Ion Tank GeLi Detector Chamber Effluent Eberline Vault Off-Gas 10-3 to 40 - 1250 Monthly Grab Sample Discharge, Storage 102 Analyzed on 2 del HP270G Tank GeLi Detector Discharge Near WI'B Tracerlab Stack Base of 10-7 to 60 - 2000 Each Re-Grab Sample Gas Model FD-5B Stack 10-1 fueling Analyzed on NAI Crystal Samples GeLi Detector 115 ft.
up Stack Discharge Eberline High Base of 10-1 to 40 - 1250 Each Re-Grab Sample Range Stack Stack 105 fueling Analyzed on Noble Gas Samples GeLi Detector Model HP270G 115 ft.
up Stack Discharge Each monitor is shielded with lead to reduce the background radiation.
Additional shielding will be added during the present outage to provide background dose protection.
All the monitors described above read out in the Control Room.
They are powered from the Turbine Building 120 Volt Regulated Bus which is fed from the Emergency Battery Bank on loss of all AC power.
EPP-21 describes procedures to be used for minimizing occupational exposures during emergencies, the methods to convert instrument readings to release rates in conjunction with the latest calibration and where to report the results.
EPP-6 describes the Emergency Control Team operations and IIcalth and Safety Procedure (HSP-02.7) describes the calibration procedure.
NRC REQUEST 2.1.8.b(2) - (Cont'd)
Radiciodine and Particulate effluents are sampled continusouly with a Tracerlab MD-4B NA-I Beta Crystal and read out in the Con-trol Room.
In addition, a sample is obtained on glass fiber and charcoa) cartridge filters for analysis on a GeLi detector.
HSP-13.6 and EPP-21 describe the analysis during interfering back-ground radiation levels.
These samplers are located at the base of the stack and sample 115 feet up the stack via an isokinetic nozzle.
Multiple GeLi gamma analyzers are available on site at four locations which are accessible during emergencies.
NRC REQUEST:
2.1. 8. c Describe the equipment in the TSC and the control room to monitor the radiciodine airborno concentration.
OPC RESPONSE:
The following equipment is present in the TSC for monitoring:
Technical Associates frisker with audible rate indication Staplex Hi-Volume Air Sampler Staplex Low-Volume Battery Powered Air Sampler
. Victoreen G-M Survey Meter
. NMC Internal Proportional Counter, alpha and beta-gamma
. Nuclear Data 512 Channel NaI Gamma Analyzer.
The following will be added as it becomes available:
Eberline RAS-1 Particulate and Charcoal Cartridge Air Sampler Nuclear Data 6600 or Northern Scientific Multi-Channel Ge-Li Gamma Analyzer.
The following equipment is located in the control Room for monitoring:
Eberline RAS-1 Particulate and Charcoal Cartridge Air Sampler i
Eberline Mini-Scaler, MS-1, with single Channel Analysis Capabilities for Air Particulate and Radiciodine Counting.
NRC REQUEST:
2.2.1.b(1)
Describe the STA short term program, including the manning requiremente, qualification and the training being given.
DPC RESPONSE:
The Shift Technical Advisor (STA) program at the LACBWR plant consists of nine individuals who rotate the duty as Shift Technical Advisor.
The rotation is on a twenty-four hour shift basis ar.d sleeping quarters have been provided on site.
The Shift Technical Advisor on duty is not permitted to leave the site unless relieved
NRC REQUEST 2. 2.1. b (1) - (Cont'd) of this duty by another qualified individual.
The individuals assigned to the position of Shift Technical Advisor are graduate engineers and department supervisors with long-time facility experience.
The engineers average 6 years of assignment at the LACBWR site and the non-engineering average 14 years of experience at LACBWR.
The non-licensed individuals in the Shift Technical Advisor group will participate during the 1980 operator requalifi-cation program lectures for emergency system, routine and emergency procedures and specific operating characteristics.
Additionally, each Shift Technical Advisor will review a special lecture on transient conditions from a consultant involved in the design analysis for emergency core cooling systems.
This lecture will especially emphasize expert plant response to small break loss of coolant accidents and instrument response to transients.
NRC REQUEST:
- 2. 2.1. b (2)
Describe the program for the long term for the STC, including minimum qualification and the training that vill be provided by 1/1/81.
The long term Shift Technical Advisor program will continue to utilize the existing individuals which have the training and exper-ience to be effective.
Additionally, new professionals being added to the LACBWR staff will be trained in the facility and then assigned Shift Technical Advisor duties.
Prior to this assignment, they will receive the licensed operator qualification training in emergency system, routine and emergency procedures and specific operating characteristics.
They will also receive the transient condition training.
NRC REQUEST:
- 2. 2.1. b (3)
Describe the organization that vill perform the oper-ational assessment function for both long and short term requirements.
Include a description of areas to be covered by this function.
DPC RESPONSE:
The Shift Technical Advisors perform the operational assessment function and will continue to do so in the long term.
The operational assessment function consists of a review of all licensee event reports, incident reports, and maintenance requests.
It also includes a review of off-site operational experiences at other plants through the Nuclear Regulatory Commission's reporting of Licensee Events sorted by type.
The Shift Technical Advisor also reviews the Shift Supervisor's Log.
These reviews'are documented.
l
NRC REQUEST:
2.2.1.c Revise the turnover procedures to include a method for conveying the status of vork on the plant being performcd by operators, etc.
Describe the method to the NRC.
DPC RESPONSE:
Shift and Relief Turnover Procedure.
LACBWR's procedure on shift transition requires each oncoming Plant Operator and Senior Plant Operator to read and initial the Plant Operator's Log Book prior to assuming their respective operator duties.
If any critical parameters are not within allowable limits when the Plant Operators take their periodic station readings for the station logs, those parameters are recorded in the Plant Operator's Log Book.
The meters for critical plant parameters are marked to t
show the allowable limits.
Therefore, when the oncoming shift reads and initials the Plant Operator's Log Book, they are made aware of any critical plant parameters that are not within allowable limits.
The outgoing Shift Supervisor is required by the procedure to review and sign all operating log sheets utilized during his shif t for the plant operation and to review the Control Room Log for accuracy and completeness.
The shift transition procedure requires that a checklist on avail-ability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents be completed during shift turnover.
The checklist specifies what is to be checked and its acceptable status.
It also provides for listing of all equipment that is in a degraded mode of operation permitted by Technical Specifications, the length of time it is permitted to be within the degraded mode, and the length of time it has spent in the degraded mode.
t The shift transition procedure also requires the outgoing operators to pass on the following information, as a minimum, to their reliefs:
1)
Status of Reactor Plant a)
Power level b)
Steam flow c)
Forced circulating pump speed /flou d)
Reactor auxiliary systems in service e)
Any abnormal, off-normal, or non-routine condition or evolution in progress f)
Any waste collection process in progress g)
Equipment out-of-service or out-of-commission.
NRC REQUEST 2.2.1.c - (Cont'd) 2)
Status of Generator / Turbine Plant a)
Generator load b)
Auxiliary systems in service c)
Any abnormal, off-normal, or non-routine evolutions in progress d)
Any equipment out-of-service or out-of-commission.
3)
Auxiliary Operator Turnover a)
Systems tagged out or cleared b)
Any evolutions or procedures in progress c)
Status of waste water system d)
Major events since oncoming crew last had watch.
Also, at the beginning of each shift, equipment out of service, keyswitch interlocks in bypass, and abnormal conditions are re-corded in the aforementioned Plant Operator's Log Book.
The effectiveness of the shift transition procedure is audited by the Quality Assurance Department as part of their periodic Operations Audit.
NRC REQUEST:
- 2. 2. 2. b (1)
Develop procedures to provide a vorkable method which vill allow you to get plant data from the Control Room to the TSC nou.
Describe the method to the NRC.
DPC RESPONSE:
ACP-2.7, " Technical Support Center", is being revised to include provision for the following:
The Emergency Control Director shall appoint a staff member to obtain desired plant data from the Control Room via the plant PABX system or walkie-talkies.
If radiation levels permit, subject to the Health & Safety Supervisor's concurrence, the Emergency Control Director shall appoint a staff member to make copies of Control Room charts which are needed by personnel staffing the TSC and bring the copies to the TSC.
Control Room charts shall not be removed from the plant building.
NRC REQUEST:
2.2.2.b(2)
Expand on your previous proposal of the long term TSC, specifically those concerning the habitability and obtaining plant da-ta from the control room to the TSC.
DPC RESPONSE:
The Technical Support Center ventilation will be modified to utilize a filtered air supply.
Shielding analysis reviews are being performed to determine the expected levels of radiation in the area following an accident.
Permanent monitoring capa-bility will be installed in the room.
The transmission of various reactor and site parameters from the control room will be accomplished by a computer.
The data points ~.will be continuously scanned.
Whenever a predetermined parameter (such as reactor water level) reaches a setpoint, a line printer in the Technical Support Center will start reading the parameters every 60 seconds..The data points will include primary system condition, core cooling, containment building conditions and site meteorological conditions.
Program capa-bility for additional transmission of data will be included.
NRC REQUEST:
2.2.2.c (1)
Establish OSC near the control room for the essential personnel and describe its location.
Revice your procedures accordingly.
(2)
Describe the use of emergency assembly points as part of the function of OSC.
DPC RESPONSE:
During a Class A Evacuation, all personnel other than the duty operating crew, the Shift Technical Advisor and the on-site Health Physics Technician report to the activated emergency assembly point.
Also, any personnel called in reports initially to the activated emergency assembly point.
The purpose of having all non-duty personnel reporting to the activated emergency assembly point is to minimize excess dose to personnel before radiation levels can be satisfactorily determined.
Thus, the activated emergency assembly point serves as the On-Site Operationa.1 Support Center initially.
After radiation surveys have been conducted,-the Emergency Control Director will send operations support personnel as needed to the Operational Support located in the Operations Supervisor's Office, down the hall from the Control Room, with the concurrence of the Health and Safety Supervisor.
Communication with the Control Room can be conducted from the emergency assembly points and the Oper-ations Supervisor's Office.
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