ML19256D999
| ML19256D999 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/19/1979 |
| From: | Mcgowan G BALTIMORE GAS & ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.E.4.2, TASK-TM NUDOCS 7910250269 | |
| Download: ML19256D999 (14) | |
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t BALTIMORE GAS AN D ELECTRIC COMPANY GAS AN D ELECTRIC BUILDING B ALTi M O R E, MARYLAN D 212 03 Aarwua E. Lunovatt,Ja. October 19, 1979 v.c c p. ..on.o ss u Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cc= mission Washington, D. C. 20555 Attn: Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors
Subject:
Calvert Cliffs Nuclear Power Plant Units Nos. 1 & 2, Dockets Nos. 50-317 & 50-318 Follow-up Actions Resulting from TMI-2 Incident (Lessons Learned)
Gentlemen:
This letter and its enclosures comprise our response to your September 13, 1979, letter on follovup actions resulting from the TMI-2 incident. Enclosure 1 in our response to NUREG 0578 as modified by your letter; enclosure 2 is our response to the "Near Term Requirements for Improving Emergency Preparedness", enclosures 7 and 8 of your letter.
We have already begun to make every effort to meet all of these requirements as they apply to our plant and vill continue to do so within the extremely short time periods allowed. However, we feel that it is counterproductive to establish ecmpletion dates that obviously cannot be met. We have therefore developed preliminary schedules for these items which, although more realistic then those provided by your Staff, must still be considered very cptimistic. Where necessary and possible, we have identified interim =easures to be taken if action is not complete by your required dates. Of course, completion dates for many items are impacted by some factors largely beyond our control, sitch as material deliveries ,
manpower availability, inputs from govern =ent agencies , or reviews by NRC.
When each NUREG 0578 item nears completion and is ready for installation, ve must emphasize that it is not vise to require a specific shutdown to accomplish that specific change. Each shutdown and startup takes the plant through conditions that are less stable than continuous power operation, and adds additional thermal cycles to plant components.
Total personnel exposures are likely to be greater during short specific-purpose shutdowns than if the work is done during an extended planned outage. We are therefore planning to i=plement those ite=s which require plant shutdown during currently scheduled refueling outages, which are in spring 1980 for Unit 1, and fall 1979 and fall 1980 for Unit 2.
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Mr. D. G. Eisetut October 19, 1979 As our plans progress and become more firm, we vill keep you informed of any significant changes that develop in our con:mitments and schedules described in the enclosures.
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Enclosures G. V. McGowan for A. E. Lundvall, Jr.
cc: J. A. Biddison, Esquire G. F. Trowbridge, Esquire Mr. E. L. Conner, Jr. - NRC 1210 141
Page 1 Enclosure 1 Response to NUREG 0578 and Enclosure 6 of 9/13 NRC Letter 2.1.1 _E=eraency Power Sunnly Reauirements Pressurizer proportional heaters are currently supplied from an emergency power supply (diesel generators). Functional requirements for the total amount of heater capacity required and timing requirements following an accident with loss of offsite power are being developed by the CE Owners Group, to be completed for implementation review by January 1980. The extent of final modifications to provide emergency power to any additional heaters is totally dependent upon the outcome of CE's work. However, in the meantime, we are in preliminary design stages of determining alternative means of temporary hookup of a bank of backup heaters to an emergency power supply. We believe we can accomplish this design, procurement of te=porary cabling, and procedures by Januar/1980.
The motive components of the power operated relief valves (PORV's) and the PORV block valves are supplied from safety related h80v motor control centers which have a diesel backup. The control components of the PORV's and PORV block valves are supplied from either the same safety related 480v motor control centers or from safety reltted 125v de battery buses.
TVo of the pressu.-izer level instruments for each unit are povered from the vital de buses and the third is povered from offsite AC power with diesel backup.
We vill review the motive and control power interfaces with the emergency buses to determine if these interfaces satisfy safety-grade requirements.
This review vill be completed by January 1, 1980 and steps necessary to upgrade components vill be started as identified.
2.1.2 Relief and Safety Valve Testing A program for testing power operated relief valves (PORVs) and safety valves (SVs) used for primary system pressure control under design bases operating conditions is being developed by the CE Owners Group. This program includes definition of test conditions and qualification requirements for all specified valves in operating reactors designed by Combustion Engineering.
The results of this program vill be made available to the generic efforts being undertaken by the industry (through for example the Electric Power Research Institute, EPRI, and the Nuclear Safety Analysis Center, NSAC) no later than January 1, 1980. These results vill also be available for discus-sions with the NRC staff to establish generic resolutions no later than January 1, 1980. Calvert Cliffs vill comuly with the schedule for completion of the test program which is agreed to during these discussions.
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2.1.3a Direct Valve Position Indication The direct position indication of the two PORV's and the two SV's per unit vill be accomplished with an acoustic monitoring system. The system vill be safety grade without redundant sensors and vill be alamed with indica-tion in the main control room. The design vill be completed for review by January 1, 1980. Installation vill be accomplished during the spring 1980 outage for Unit No. 1 and fall 1980 outage for Unit No. 2. Presently there are various indicators to help determine the status of the valves, such as power indication to the PORV solenoids, and quench tank pressure, temper-ature, and level indication, all of which is located on the front of the same control panel in the control room. In addition, there are thermocouples located in the two discharge lines. These indications provide adequate means of determining valve position during the period of time required to design the new system, procure equipment and cable, and complete installation.
Plant procedures previously in effect to provide guidance to operators in use of these indications to determine valve status were reviewed in detail in light of the TMI incident, and were upgraded to include any necessary improvements.
2.1.3.b Instrumentation for Detection of Inadeouate core cooling Procedures and Additional Instrumentation To the extent possi!O
- vithin the framework of existing analyses, procedures have been upgraded in response to IE Bulletin 79-06B to aid operators in detection of inadequate core cooling and to assure appropriate actions are taken. Additional procedures to be used by an operator to recognize inadequate core cooling vill be developed based on analyses being performed as required by Ites 2.19, Transient & Accident Analysis, Analysis of Inadequate Core Cooling. The guidelines for the procedures are being developed by t!'e CE Owners Group and vill be available for discussions with the NRC staff to establish generic resolutions no later than January 1,1980.
If the analyses or the guidelines ir iicate the need for the design of new instrumentation, the design of such instrumentation vill be made available for discussions with the NRC staff to establish generic resolutions and appropriate schedules.
Subcooled Margin Monitor A conceptual design for the subcooled margin monitor is expected to be completed by January 1,1980, with installation of the monitor in the spring of 1980 for Unit 1 and in the fall of 1980 for Unit 2. The limited range on existing process temperature signals, e.g., hot leg temperature, and heavy signal loop loading are problems currently under engineering review.
Resolution of these range problems will have a major impact on final design.
As an interim =easure, we have begun development of a program addition to the plant computers which will provide ecsparable information until the pernanent installation can be completed. We intend to have this program operable by January, 1980.
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. .. . , - - - - - - - - - - - Page 3 2.1.h Diverse Con _tainment Isolation The containment isolation system design for Calvert Cliffs is being modified so that there vill be diversity in the parameters sensed for initia-tion of containment isolation. This change vill be completed by January 1, 1980, for Unit 2, and during the spring 1980 outage for Unit 1.
A preliminary study of all fluid lines penetrating the containment for the purpose of reconsidering the definition of essential and non-essential systems has been completed. A generic review of all systems penetrating the cont .inment on all opersting plants with a CE designed NSSS is being conducted by the CE Owners G*oup. This review will produce generic criteria for 'he definition of essential systems, identification of all such systems, and specification of the bases for selection of each system. Criteria are also being developed for selective unisolation of non-essential systems which may be beneficial. The results of this review vill be submitted for implemen-tation review by January 1, 1980.
The prelimina.y study used, as its basis for selecting essential systems, the criteria that essential fluid systems vill be those that are actively required during the early stages of an accident to control and mitigate the consequences of an accident such that exposure to off-site individuals is not in excess of the limits specified in 10 CFR, Part 20.
The essential fluid systems into containment are:
- 1. High and low pressure safety injection;
- 2. Containment spray water;
- 3. Service water to containment structure cooling units; and
- h. Auxiliary Feedvater (after approximately 15 minutes only)
The essential fluid systems out of containment are:
- 1. Containment structure sump recirculation to safety injection pumps;
- 2. Service water outlet;
- 3. Main steam lines; and
- h. Containment pressure instrument lines.
These essential fluid systems are not automatically isolated by contain=ent isolation which is initiated by the safety injection actuation system (CIS/SIAS), nor by the containment isolation actuation signal (CIS).
The re=aining fluid lines penetrating the containment are considered non-essential. These penetrations are either normally locked closed, close on CIS/SIAS, close on CIS, close on SGIS, or close on containment high radiation signal (CR3). FSAR Table 5-2 describes all of the fluid system containment penetrations and their valve arrangements.
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To effect a design change to the control schemes for the containment automatie isolation valves, so that manual resetting of the isolation lignal vill not result in the automatic reopening of the contain=ent isola-tion valves, would require additional relaying in the control scheme.
For the reasons described below, this additional relaying vill not substan-tially improve the safe operations of the plant nor reduce the ritt of radiation exposures to off-site personnel. The additional relaying is viewed as actually decreasing the reliability of the containment isolation system by adding additional, unnecessary hardware that has the potential for malfunctioning and negating control rocm operator control of the isolation valves and/or disrupting nc mal operation and creating unnecessary challenges to the safety systems.
The Calvert Cliffs containment isolation system (part of ESFAS) is not automaticall/ reset when the initiating parameters return within bounds; the systam does not allow blocking nor overriding the isolation signals, neither manually nor via any process control signal; and, administrative procedures and controls have been established to assure the continued operability of the isolation system.
These features make it unnecessary to modify the automatic isolation system.
These features assure that reopening of the contain=ent isolation valves vill require deliberate operator action, i.e., unlocking the ESFAS cabinet doors, manually resetting the individual trip logic modules, and then returning the individual isolation valve handsvitches to the open position.
2.1.5 Post-Accident Hydrocen Control The Calvert Cliffs post-accident hydrogen control system is described in FSAR Section 6.8. This system currently meets short and long tem requirements of the NUREG.
2.1.6a Integrity of Systems Containine Radioactivity BG&E is currently evalut. ting the feasibility of limiting the number of systems which will contain radioactive fluids following an accident, which would include an evaluation of effecting cooldovn of the reactor coolant system to <2000 F using the steam generators. Assuming that this approach is followed, it is likely that the January 1980 deadline for implementation of an i= mediate leak reduction program and a preventative maintenace program for those systems can be met. If it becomes necessary to verify leaktight integrity of additional systems, a correspondingly longer time frame for program implementatior. vill be required. The actual date that implementation can be completed vill depend upon the number of systems which must be tested.
It is anticipated that our evaluation vill be completed and the number of systems to be tested will be determined by November 15, 1979 At that time a fim co=mitment date, which vill be no later than January 1981, can be made.
2.2.6b Plant Shieldine Review A preliminary design review of plant shielding of spaces for post-accident operations is being conducted. By January 1, 1980, the doses to the control room from the systems required for post-accident recirculation vill be 1210 145
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2.1.6b Plant Shielding Review (continued) identified and a conceptual shielding design provided if found to be necessary. This design reviev vill be further expanded to include all the lines carrying radioactive material and all the areas which must be accessed for post-accident operations. By January 1981, te=porary shielding, as a minimum, vill be provided in all essential areas.
2.1.7a Automatic Infation of Auxiliary Feed Automatic initiation of auxiliary feedvater is considered to be unnecessary for Calvert Cliffs for the following reasons: (1) immediate actions are required by the reactor trip procedure to verify feedvater flow status; (2) there is complete control of the auxiliary feedvater system from the main control board; (3) there is approximately fifteen minutes available before auxiliary feedvater is required; and (h) past experience with recovery from feedvater system problems indicates no need for automation of the auxiliary feedvater system. A program to calculate and document the time available before dryout of the Calvert Cliffs steam generators following total loss of feedvater flow from full power conditions is being conducted by the CE Owners Group. The results of this program vill be submitted for proposal review by the NRC staff no later than January 1, 1980. Following that reviev, a decision vill be made concerning the need for installation of equipment to provide automatic initiation of auxiliary feedvater in Calvert Cliffs.
An engineering package for automatic start of the auxiliary feed pumps on loss of main feed was developed, but we believe its value is negligible and do not now intend to implement it. Current design and operating pro-cedures call for manual initiation and control of auxiliary feed on loss of main feed. Following any reactor trip, auxiliary feed comes under the direct control of a dedicated operator. In order to implement a fully automated start and control scheme for the auxiliary feed system, substan-tial re-analysis is required in light of concerns associated with a steam line break accident. Insufficient time was available under NUREG 0578 for complete analysis and adequate review, and the design package was therefore limited in scope to automatic start of the auxiliary feed pumps; the operator vould retain control of the auxiliary feed regulating valves. Since operator action vould be required on loss of main feed with or without auto =atic start of the auxiliary feed pumps, the implementation of that feature vould not provide any significant improvement in system reliability.
Investigation is continuing on alternative design modifications of the auxiliary feed system, including automatic start and control of the total system, and other physical modifications. System design modifications, if any, vill be determined after review of inputs from the NRC Bulletins &
Orders Group, the NSSS Vendor, and the Architect-Engineer.
2.1.7b Auxiliary Feed Flov Indication A safety grade auxiliary feedvater flow indication for Unit 2 vill be in service by January 1, 1980. Currently, auxiliary feed flow indication is in service on Unit 1; it vill be upgraded to safety grade during the spring 1980 outage.
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Page 6 2.1.oa Imoroved Post-Accident Samuling Canability A preliminary design and operational review of both the reactor coolant and containment atmosphere sampling facilities vill be done by January, 1980, to determine our capability to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/h Rems to the whole body or extremities respectively. Results of the preliminary review and an outline of resulting procedures and any plant modifications will be available by January, 1980.
A preliminary design and operational review of the radiological spectrum analysis facilitiec vill be performed to determine our capability to promptly quantify (less than 2 hovrs) certain radioisotopes that are indicators of the degree of core damage. The review vill consider radiation effects as described in the NUREG. Results of the preliminary review and an outline of resulting procedures and any plant modifications vill be available by January, 1980.
Our current procedures regarding the analysis of boran and chloride vill be reviewed by January, 1980, to determine if changes should be made to allow analysis assuming a highly radioactive initial sample while maintaining exposures within the limits described above. The procedural review vill encompass the feasibility of completing the analysis for boron within an hour and the chloride within c shift.
2.1.8b Increased Range of Radiation Monitors Wide range noble gas monitors with associated indicators and recorders vill be installed at the appropriate release points by January 1, 1981. These monitors vill be povered from the vital instrument buses and vill be seis-mically qualified. Our present monitors presently measure up to 1 p Ci/cc vith a release rate of 45 ci/sec, which is greater than those at BfI.
A review of our main vent instrumentation measurement capability for noble gases, radiciodines and particulates vill be performed as well as our procedures for sampling and analysis of grab samples. A method vill be developed by January, 1980, to estimate release of noble gases, radiciodines and particulates in the event the main vent instrumentation goes off see.le.
Redundant high range radiation monitors for the containments with associated indica 6 tors and recorders vill be installed by January 1, 1981. They vill be seismically qualified and povered from a vital instrument bus.
2.1.8e Innroved In-Plant Iodine Instrumentatien Even though current iodine =easurements are perfor=ei using ga=ma energy spectrum analysis, a review of our techniques vill be performed by January, 1980, to determine if improvements of equipment or technique are necessary to assure accuracy under accident conditions. Both in-plant techniques and offsite techniques vill be included in the review.
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Page 7 2.1.8c Imoroved In-Plant Iodine Instrumentation (continued)
Although the NUREG 0578 position on this item appears to address the necessity for use of respiratory equipment, it should be pointed out that there is currently no NRC approved respirator cartridge for iodine removal.
Current regulations (10CFR 20.103 and Regulatory Guide 8.15) only allow a protection factor equal to 1, and therefore intakes of radioactive materials (iodine) are not " legally" reduced by virtue of the use of an unapproved respirator cartridge. It is suggested the NRC review its position in this area so that respirator use can be properly prescribed to keep intakes ALARA.
2.1.9 Transient and Accident Analysis The response to Transient and Accident Analysis require =ents is being developed by the CE Owners Group in conjunction with generic resolution meetings with the NRC Bulletins and Orders Task Force. These responses vill be submitted on the schedule agreed to by that Task Force and the CE Ovners Group and vill be referenced for specific application to Calvert Cliffs.
AMitional Equinment:
Containment Pressure Indication Continuous indication in the control. room of containment pressure in the range of (-)5 to 150 psig vill be installed by January 1,1981. The trans-mitters vill meet the requirements of Regulatory Guide 1.97 Revision 1.
The new revision of Regulatory Guide 1.97 vill be reviewed for appropro-priateness when received. Presently there is the capacity to measure from
(-)h to 60 psig with indication in the main control room.
Containment Water Level Indication Presently two level transmitters per unit exist to measure the entire sump level with indication and alarm in the main control room. These transmitters are qualified for a LOCA environment. A vide range level transmitter, qualified for a LOCA environment, vill be installed in both containments with indication and alarm in the =ain control room by January 1, 1981. The transmitters vill meet the requirements of Regulatory Guide 1 97 Revision 1.
The new revision cf Regulatory Guide 1.97 vill be reviewed for appropriateness when received.
Containment Hydrogen Indication There are redundant hydrogen recombiners in each containment. Each one is designed to maintain the hydrogen concentration to less then h volume percent. The recombiners are designed to operate under LOCA conditions, are operated from the s. trol room, and are povered by separate, redundant, emergency sources. The hydrogen analyzer, with a 0 h% range, is located in the auxiliary building with an alarm in the main control room. Because of this system, a 0-10% H 2 e neentration range for an analyzer is considered to be not needed. The results of an investigation into the accessibility to the analyzer room after a LOCA vill determine whether control and indica-tion of the analyzer is necessary in the main control room.
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2.1.9 Transient and Accident Analysis (continued)
RCS Venting The functional requirements and conceptual design of a system for remote venting of the RCS are being developed by the CE Owners Group. These requirements and a conceptual design vill be available for discussions with the NRC staff no later than January 1, 1980, to establish generic resolutions. A decision concerning 1istallation of an RCS venting system in Calvert Cliffs vill be made following these discussions.
2.2.la Shift Sunervisor Responsibilities The requirements of the NUREG position vill be instituted by January 1, 1980.
2.2.lb Shift Technical Advisor The present plans to satisfy the NUREG position regarding the Shift Technical Advisor (STA) are based on discussions with the NRC staff held on this subject at the topical meeting in Bethesda, MD on October 12, 1979 Even though our plans do not meet the literal requirements of the NURE position, we believe ve vill meet the intent of the NUR E as expressed at the October 12 meeting. Specifically, ve intent to separately institute measures to satisfy the two major goals of the NUREG discussion regarding the duties and functions of the STA, as set forth below:
Onerating Exnerience Assessment Before January 1,1980, ve intend to establish a standing committee to regularly review the operating experience at the Calvert Cliffs plant and at other plants of like design. This comittee vill be primarily composed of staff engineers located on-site and will be augmented as necessary by engineers from our Engineering Department (located in Baltimore). The pi"7ose of the reviews conducted by the committee vill be to provide an independent monitoring of the safety of plant operations and to provide, if necessary, a counter-balanced perspective to the needs of the co = ercial apsects of plant operation. Additionally, by the review of the operating experiences of other plants of like design, valuable input may be derived which could improve the safety of the Calvert Cliffs plant.
The Chairman of this comittee vill be located onsite acd vill make reports directly to the Offsite Safety Review Comittee. The ons.'.te comittee vill not be staffed by members of the plant operations unit, although such persons may be called upon by the chair =an to provide input as necessary to the co=mittee's deliberations.
In order to assure that the facets of the review process which are pertinent to operator and technician training are propcrly factored into the training program, the plant training coordinator vill be a permanent member of the comittee.
Presently, many of the aspects of such a review process are carried out by diverse organizations both onsite and offsite. *his being true, the actual time spent presently in conducting such reviews is not formally documented.
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Page 9 2.2.lb Shift Technical Advisor (continued) reviews, future de= ands on the personnel involved is not presently clear.
Should the workload of performing these reviews become overly burdensome to present staff personnel, the staff complement vill be increased to provide an adequate number of engineers to properly conduct the committee reviews.
Accident Assessment and Resnonse In order to provide the dedicated accident assessment and response function set forth by NUREG 0578, we plan to utilize the present complement of Senior Operator License (SOL) persons assigned to the shift operating organ-ization. In doing so, we believe it is of paramount importance to assure that at all times a trained, qualified individual who is current in the operational status of the units is immediately available to the Control Room to make such assessments. This individual can have no " hands-on" respon-sibilities to perform during the response phase of an accident condition.
In order to assure that the individual performing this function remains detached from " hands-on" duties, we plan to institute the following measures by January 1, 1980.
- 1. Each unit which is in an operational mode above Cold Shutdown vill always have two Operator License (OL) persons assigned to the unit.
Since the units share a combined Control Room , during normal operation this vill provide four 0.L. persons in the Control Room. Administrative procedures vill establish the manning level such that only one of the four can be absent at any one time and then only for brief periods (approximately ten minutes). Since all of our Control Room Operators are licensed on both units, this vill insure that each operating unit vill always have two full-time, dedicated " hands-on" operators in the event of an accident situation. A majority of the time, the " extra" operator from the non-affected unit vill also be available to assist in " hands-on" functions for the affected unit.
- 2. Since our Technical Specifications require the continual presence of two SOL operators onsite,the implementation of OL manning level described above vill assure that one of the SOL operators vill always be available to perform the desired accident assessment function. We plan to assign this responsibility to the Senior Control Room Operator (SCRO), thus allowing the Shift Supervisor to perform the " Command and Control" functions associated with an accident response. The specifi:
responsibility of the SCR0 in performing accident assessment functions vill be included in a management procedure and emphasized in the training of these personnel.
As discussed at the October 12 topical meeting, an accelerated training course vill be provided for SOL personnel to enhance their ability to perform the accident assessment functions. Our training organization is presently investigating the scope and source of such training. It is expected that the training vill include such general subjects as reactor physics, reactor thermodynamics, fluid mechanics, heat transfer and reactor control theory.
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Page 10 2.2.lb Shift Technical Advisor (continued) conditions vill also be included in the course of instruction. We plan to conduct this training on a college level beginning as early in 1980 as possible. Sources of such training may include universities, junior colleges, various consultants and the NSSS vendor. In order to support such an intensive training program, we intend to divide our present complement of twelve SOL operators into three groups such that one-third are in training full time while the others provide the necessary shift coverage. Should the demands of shift coverage and training dictate, serious consideration vill be given to increasing this complement to support such a program.
2.2.lc Shift and Relief Turnover procedures The requirements set forth by the NUR E position vill be instituted by January 1, 1980. In our initial review of the concepts set forth therein, we feel that most of the elements of this position are already in place, even though they are not in a for=alized checklist. In some cases, to provide such a checklist vould be a duplication of already existing mechanisms; for example, the logging of plant parameters and limits on the checklist would be a duplication of our existing Control Room Log. In cases where such duplication vould exist, the checklist may be used as a mechanism of documentation, rather than instituting a duplicative procedure.
2.2.2a Control Room Access The requirements set forth by the NURE position vill be instituted by January 1, 1980.
2.2.2b Technical Sunport Center In order to implement this position, we intend to establish an onsite Technical Support Center (TSC) in the 55' elevation of the Log and Test Instrument Room, shown on FSAR figures 1-8 and 1-12. This room is approx-imately 23 feet by h0 feet and is included in the ventilation rytten provided for the Control Room / Cable Spreading Room (described in Pfi3 Section 9.8.2.3). Additionally, this room is within the Seismic I boundary of the Auxiliary Building. As can be seen from the FSAR figures, the room can be entered directly from the Turbine Building, and also is connected to the Control Room via a single door on the k5' elevation.
The present communications capabilities from this room includes three conventional telephones and a plant page system. Upgrading of the entire plant telephone system is presently in progress, and expansion of the phone system for the TSC vill be included in this upgrading. The final plans for this upgrading are expected to be completed by the end of this year.
A plant " Record Retention and Retrieval Center" is presently located in the northwest corner of the Control Room. This center provides controlled copies of all the documents listed in the NUREG position. Since this center is presently located in the rear of the Unit 1 control panels, access from the Log and Test Instrument Room to the center can be made without intruding into the front-panel area of the control panels. Consideration is presently being given to relocating the center to the Log and Test Instru=ent Room.
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2.2.2b Technical Suvuort Center (continued)
More detailed information regarding upgrading the TSC to meet the yet-to-be-formalized NRC requirements for instrumentation and data-links vill be provided by January 1, 1980.
2.2.2e Onerational Sunvert Center The requirements set forth by the IUREG position vill be instituted by January 1, 1980.
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Enclosure 2 Response to Enclosure 8 of o/13 NRC Letter Near Term Emereeney Prenaredness Introvements Item 1. Emergency Plan Our Site E=ergency Plan for Calvert Cliffs vill be upgraded to Regulatory Guide 1.101 criteria with Emergency Action Levels discursed in the draft NURE -0610 by mid-1980.
Ite:t 2. Instrumentation to Follow the Course of rin Accident See response to NUREG 0578 Section 2.1.8.
Item 3. Emergency overations Center The Calvert County Civil Defense Emergency Control Center in Prince Frederick performs the function of the emergency operations center for Federal, State and Local personnel. The center is currently established with suitable com=unications to the plant.
Item h. Offsite Monitoring Improved offsite thermoluminescent dosimeter monitoring capabilities have been provided and vill be documented in a future revision to our Environ-mental Technical Specifications by mid-1980.
Item 5. State and Local Plans The State of Maryland and local agencies' Radiological Ehergency Response Plans relative to the support of Calvert Cliffs are in process of revision to include addressing the elements described in NUREG-75-lli, and it is expected that these plans vill receive NRC concurrence by mid-1980. These activities, of course, are largely beyond our control. Future revisions vill be performed, when appropriate, to include upgraded NRC criteria.
Item 6. Test Exercises A test of our upgraded Site E=ergency Plan and the State of Mazyland's upgraded Radiological E=ergency Response Plans vill be accomplished by mid-1980. A joint test exercise vill be performed within a 5 year period subsequent to the mid-1980 test.
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