ML19259D655
| ML19259D655 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/20/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7910250355 | |
| Download: ML19259D655 (28) | |
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l WISCONSIN Electnc eona couraur 231 WEST MICHIGAN, MILWAUKEE, WISCONSIN 53201 October 20, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.
S. NUCLEAR REGULATORY COMMISSION Washington, D.
C.
2u555 Attention:
Mr. Darrell R. Eisenhut, Acting Director Division of Operating Reactors Gentlemen:
DOCKETS 50-266 AND 50-301 IMPLEMENTATION OF NUREG-0578 POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 Your letter to all operating nuclear power plants dated September 13, 1979 discussed follow-up actions resulting from the NRC Staff review of the Three Mile Island (TMI) Unit 2 accident.
Specifically, the letter directed that we commit to implement the actions contained in NUREG-0578, as modified and supplemented in the letter, in accordance with the schedule provided.
Our responses regarding these items are discussed point by point in the attachment to this letter.
In addition to our responsos to your September 13 letter, we hava unclosed for your information a copy of Wisconsin Ulectric Power Company's TMI Accident Review Task Force report, which was ccmpiled as a part of our review of the TMI accident.
Our Task Force was formed shortly after the TMI accident to review the events which occurred in detail to determine whether any equipment, design, system, operating procedure, maintenance program, personnel qualification, or training procedures should be modified or changed to assure that continued operation of the Point Beach Nuclear Plant does not present an undue hazard to public health and safety or to the health and safety of the employes who are charged with its operation and maintenance.
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Mr. Harold R. Denton October 20, 1979 You will note ths' many of the items considered by our Task Force are related to items addressed in NUREG-0578.
In some cases, the attached responses to NUREG-0578 include references to this Task Force report.
Your attention is also invited to the description of design differences between Point Beach and TMI, which we believe significantly reduce the probability of a TMI-type accident occurring at Point Beach.
Very truly yours, f
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Exe utive Vice President Sol Burstein Attachments 1211 087 O
2.1.1 EMERGENCY POWER SUPPLY REQUIREMENTS FOR THE PRESSURI"ER HEATERS, POWER-OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN PWR'S Pressurizer Heater Power Supply The Westinghouse Owners Group has established that pressurizer heater capacity of 100 kW is necessary for proper pressure control in the establishment of natural circulation for a two-loop plant.
The Point Beach Nuclear Plant (PBNP) design has four back-up heater groups, olus a control group.
Two of the back-up heater groups, 200 kW each, are powered from the safeguards power supply; one frcm the B03, 480 volt bus and the 3D diesel; and one from the B04, 480 volt bus and the 4D diesel.
The control power for the heater breakers is also supplied from the respective vital bus control power, the "A" battery supplying the B03 bus control power, and the "B" battery supplying the B04 bus control power.
The power supply breakers and control power that tie the heater groups with the emergency buses are qualified to the same safety standards as the remainder of the vital safety-related breakers at PBNP.
The back-up heaters supplied by vital power supply are stripped by a safety injection signal in the PBNP design.
In order to restore the heaters, the safety injection signal must be reset, and the non-safeguards lockout relays must be reset (both from the main control room).
The procedures to specify the proper timing and loading of the pressurizer heaters will be completed in conjunction with the emergency procedure effort underway with the Westinghouse Owners Group; this action will be completed by January 1, 1980.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators A review of the PBNP power suppliers for pressurizer power-operated relief valves (PORV's), block valves, and level indication instrument channels has been completed.
The motive and control components of the pre.ssurizer PORV and block valves are powered frcm engineered safeguards buses, vital instrument (battery / inverter) buses, er' directly from a battery.
The power supply breakers and control power that tie the PORV and block valves with the emergency buses are qualified to the same safety standards as the remainder of the vital safety-related breakers at PBNP.
Two out of three of the pressurizer level indication instrument channels and the cold calibration channel are powered from the vital instrument buses.
The third channel is presently powered by an AC/AC motor-generator set from a non-safeguards bus.
A modification request has been initiated to provide a battery-inverter power supply, similar to the vital buses, for this channel.
The date for ccmpletion of this change will depend upon availability of equipment.
This condition does not, however, affect the operator's ability to obtain reliable pressurizer level indication using the remaining redundant channels following a loss of off-site power.
2.1.1-1
2.1.1.
(continued)
The current PBNP power supply configuration for the pressurizer heater, PORV's, block valves, and level instrumentation, therefore, provides the operator with the capability to maintain pressure control for the reactor coolant system and monite: pressurizer level when off-site power is not available.
Except for the procedure modifications, no changes are required to meet the NRC recommendations in this area.
e 1211 089 2.1.1-2
2.1.2 PERFORMANCE
ING FOR 7WR AND PWR RELIEF AND SAFETY
,.3 Wisconsin Electric Power Company is a member of the Westinghouse Owners Group.
The Westinghouse Owners Group is working in conjenction with the other PWR owners and the Electric Power hesearch Institute (EPRI) to develop a program for qualification of relief and safety valves under expected operating conditions, including solid-water and two-phase flow conditions.
The Owners Group program description will be submitted by January 1, 1980.
We understand that the NRC staff position in respect to performance testing for safety and relief valves is based principally, if not exclusively, on the failure of the power-operated relief valve at TMI to reclose.
Such a failure to reclose of a RCS relief or safety valve, of course, results in the loss of integrity of the primary coolant pressure boundary.
All plants have been analyzed for a complete range of break sizes and no license has been issued where the LOCA analysis was deemed inadequate.
We fail to see in what manner a stuck-open relief valve is different from a small break LOCA.
We believe that a relief valve testing program will require very large expenditures and diversion of scarce personnel.
It is almost certain to demonstrate that reclosing of any safety or relief valve is not dependable, a fact that the power industry has known for decades.
Relief and safety valves, however, are designed and installed for over-pressure protection, not for their dependable relosure characteristics.
Indeed, PORV's are normally provided to prevent such dependence on Code safety valves, and motor-operated block valves are installed in anticipation of their use to respond to those characteristics of PORV's.
While we are participants in the Owners Group responding to the staff concerns in this area, we believe testing programs for safety and relief valves are unnecessary.
If the NRC staff has further justification for its position in this area, we will be pleased to review it.
Our response to Items 2.1.3a and 2.19 are related to this item also.
1211 090 2.1.2-1
2.1.3 INFORMATION TO AID OPERATORS IN ACCIDENT DIAGNOSIS AND CONTROL 2.1.3.a Direct Indication of Power-Operated Relief Valves and Safety Valve Position for PWR's and BWR's The Point Beach Nuclear Plant pressurizer pcwer-operated relief valves (PORV 's) have direct, positive indication in the control room derived from stem-mounted limit switches.
The switches will be replaced with safety-grade switches by May of 1980.
Additionally, temperature indication is provided for the common PORV discharge header.
No modifications are required to meet the recommendation for direct indication of PORV position.
The control room operator can derive a positive indication of the position of the Code safety valves based on direct indications in the control room.
Individual temperature indication is provided for the discharge header of each Code safety valve.
Positive flow indication can be determined from the pressure, level, and temperature indications for the pressurizer relief tank.
Our evaluation of the need for a direct flow detection device has determined that sufficient information already exists, and no change is required (see Sections 4.2.2 and 5.3.4 of the attached Wisconsin Electric Power Company TMI Accident Review Task Force report).
The Westinghouse Owners Group, of which Wisconsin Electric Power Company is.a member, is performing transient and accident analyses and evaluations which address small break, loss-of-coolant accidents generically and independent of location.
These results, revised procedures, and implementation schedule are addressed in Item 2.1.9.
The analyses consider PORV and Code safety valve releases without the need to identify specifically the individual component condition.
2.1.3-1
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2.1.3 INFORMATION TO AID OPERATOP.S IN ACCIDENT DIAGNOSIS AND CONTROL 2.1.3.b Instrumentation for Detection of Inadecuate Core Cooling in PWR's and BWR's Procedures and Descrirtion of Existing Instrumentation The Westinghouse Owners Group, of which Wisconsin Electric Power Company is a member, is performing calculations associated with the definition and recognition of inadequate core cooling in response to NUREG-0578, Item 2.1.9, Transient and Accident Analysis. Refer to the response to Item 2.1.9 for the program scope and schedule description. The results of the program are expected to provide guidelines which can be incorporated into individual plant procedures and to establish the usefulness of existing plant instrumentation in identifying the approach to, and the existence of, inadequate core cooling. Subcooling Meter Wisco.in Electric Power Company, in its Task Force review of the accident at Three Mile Island, determined that an indication of subcooling in the primary system would be beneficial to the operator a.nd -oda a recommendation that this information should be provided to the operators at Point Beach. As an interim means-of providing an indication of subcooling, the data-logging computer for each unit will be programmed to alarm when there is less than 50*F of subcooling in the core. The average of the in-core thermocouples will be used as the reference temperature. This will be done by January 1, 1980. Other means of indicating subcooling are presently under review. It is expected that a system using existing plant sensore to calculate subcooling, separate from the data-logging computers, will be installed by July 1, 1980. Design and Installation of New Instrumentation The program described above will also provide the basis for any new instrumentation which could improve the capability of the operator to recognize the conditions which could lead to inadequate core cooling or to recognize that inadequate core cooling exists. Particular attention will be paid to the need for vessel level instrumentation and how it would be utilized by the operator. If a significant net benefit can be derived from the installation of vessel level instrumentation, disconsin Electric Power Company will provide such instrumentation for the Point Beach Nuclear Plant. A description of any new instrumentation, its functional design requirements, and a schedule for installation will be provided two months after the analysis is completed, as outlined in Item 2.1.9. 1211 092 2.1.3-2
2.1.4 CONTAINMENT ISOLATION PROVISIONS FOR PWR'S AND BWR'S The cantainment isolation system utilized at the Point Beach Nuclear Plant was described in Wisconsin Electric Power Company's response to Item 9 of Bulletin 79-06A dated April 27, 1979, and is actuated by a safaty injection signal alone or upon high containment pressure. Further evaluation will be performed which will identify all essentiel and non-essential systems, and this evaluation will be provided to the NRC by January 1, 1980. This evaluation will also identify the systems which are normally isolated during operation, the systems which are isolated upon initiation of a containment isolation signal, and those systems which are not isolated. It is expected that this additional information will demonstrate the adequacy of the present containment isolation system utilized at the Point Beach Nuclear Plant. 1211 093 ~ 2.1.4-1
2.1.5 POST-ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND BWR CONTAINMENTS 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems Appendix D to the Point Beach Nuclear Plant (PBNP) Final Facility Description and Safety Analysis Report (FFDSAR) contains the PBNP post-LOCA containment purging analysis and a description of the post-accident containment ventilation system (PACVS). For this analysis, the required flow rate of the purge system is 240 cfm, assuming the system is operated one hour per day with the activity release based upon TID-14844. The PBNP system includes the following features: a three-inch exhaust pipe from a plenum near the top of the containment structure, a manual flow control valve and manual isolation valve outside containment but upstream of the pipe's entrance into the primary auxiliary building exhaust system, a 3/4-inch sample connection originating from the top of the containment, and a two-inch return line to containment designed to be connected to the service air system in the primary auxiliary building, if required. Operation of the system is specified in EOP-llA, Post-Accident Ventilation System. A design review calculation is being performed to verify the original design with respect to the flow capacity of the purge system. Modifications will be implemented, if required, as soon as the results are completed. Analysis work will be completed prior to January 1, 1980. The PBNP post-accident purge system design is in accordance with the containment design criteria in effect at the time PBNP was licensed prior to both General Design Criterion No. 54 and No. 56. The as-built system has been reviewed and found to conform to the containment design criteria approved for PBNP. Redundant isolation is provided with two manual valves (locked closed) in series in each line. 2.l.5.c Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant The position description, as modified by the NRC letter of September 13, 1979, requires a review of procedures and bases for recombiner use. The Point Beach Nuclear Plant PACVS has the capability to accommodate an externnt recombiner for post-accident hydrogen control within a few days dter an accident. We believe, however, that the present purging analysis, as described in Appendix D to the Point Beach FFDSAR, is still the preferred, acceptable method for post-accident combustible gas control. This subject is also discussed in the Wisconsin Electric Power Company TMI Accident Review Task Force report in Section 5.23. 1211 194 2.1.5-1
2.1.6 POST-ACCIDENT CONTROL OF RADIATION IN SYSTEMS OUTSIDE CONTAT.NMENT OF PWR'S AND BWR'S 2.1.6.a Intecrity of Systems outside Containment Likely to Contain Radioactive Materials (Encineered Safety Systems and Au:<111ary Systems) for PWR's and BWR's An analysis will be made to ider.ify those systems that would, or could, contain highly radioactive fluids during a serious transient or accident. This analysis will be completed by January 1, 1980. Immediate leak reduction measures and actual leakage rate determinations will commence as soon as practicable upon completion of the analysis. A continuing leak reduction program will be established to reduce leakage to as-low-as-practical levels and will include periodic testing. It is anticipated that this effort will be incorporated into the Point Beach Nuclear Plant in-service testing program and will be performed in conjunction with testing required in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. It is also anticipated that provisions for testing of the gaseous piping systems (e.g., using helium) will need to be evaluated. Should such a system prove to be feasible, modifications to the plant piping system will have to be made. Specific methods and criteria are needed to analyze whether a new leak testing system will give the response desired. This evaluation will be completed prior to January 1, 1981. )2ll h9b 2.1.6-1
2.1.6 POST-ACCIDENT CONTROL OF RADIATION IN SYSTEMS CUTSIDE CONTAINMENT OF PWR'S AND BWR'S 2.1.6.b Design Review of Plant Shielding and Environmental Qualification of Eauipment for Spaces / Systems Which May Be Used in Post-Accident Operations Point Beach Nuclear Plant was designed for cenLinued operation with one percent fuel element defects. A review will be performed, primarily of the Auxiliary Building, to determine the radiation levels resulting in spaces around systems that may, as a result of an accident, contain more highly radioactive materials. This design review will consider the radiation levels in areas to which personnel access is required during post-accident recovery operations. Significant safety equipment components required for post-accident operations will also be reviewed to assure that undue degradation by radiation fields does not occur. This review will be accomplished with the following steps: a. Identification of the minimum number of systems which must be available for post-accident operations. b. Derivation of an appropriate and realistic source term to apply to gases or liquids contained in those systems. The source term proposed by the NRC staff (e.g., 100% of core inventory noble gases in liquid systems and 100% of core inventory noble gases in the containment atmosphere) is over-conservative and unduly restrictive. Calculation of radiation levels in' appropriate access c. spaces and spaces around pertinent safety equipment. d. Evaluation of personnel access and resulting doses and evaluation of integral radiation exposures of safety equipment. e. Provision of permanent shielding, temporary shielding, and proceric 21 controls as appropriate. This analysis is a.cajor work effort, and cannot be completed in accordance with the schedule proposed by the NRC. It is estimated that the design review may be ecmpleted by January 1, 1981, and a reasonable amount of plant modifications, if neeced, could be implemented by January 1, 1982. r 1211 096 2.1.6-2
2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWR'S 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWR's The Point Beach Nuclear Plant (PBNP) auxiliary feedwater system is a safety-grade system which provides for automatic initiation and is designed to meet single-failure criteria. The capability also exists to initiate manuall; auxiliary feedwater from the control room. Testability of the initiating signals and circuits exist to the same degree as for other safety-grade systems. The initiating signals and circuits are powered from the emergency buses. The auxiliary feedwater system design, instrumentation, actuation, and operation has been the subject of ongoing discussion with the NRC beginning with IE Bulletin 79-06B. Extensive, detailed information on the PBNP auxiliary feedwater system was provided to the NRC (Messrs. Trammell and LeFave - NRC, Asselin-Sandia Labs) in a May ll, 1979 meeting, as documented by the May 15, 1979 letter from Mr. Trammell to Wisconsin Electric Power Company. 'This and other additional information provided informally to the NRC are included in an NRC staff report to the Commission which is currently in draft form. A separate response is being prepared in reply to the September 21, 1979 letter from Mr. D. Eisenhut to Mr. Sol Burstein entitled, "NRC Requirements for Auxiliary Feedwater Systems at Point Beach." These responses will be completed before January' r,- 1980". e 1211 097 2.1.7-1
2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWR',S 2.1.7.b Auxiliarv Feedwater Flow Indication to Steam Generators for PWR's Flow orifices and local indication of flow are presently provided on the discharge of each auxiliary feedwater pump. Safeuy-grade flow transmitters will be connected to these orifices for control room indication. The channels will be routed through the protection rack through hardware similar to that for existing reactor protection system instrumentation. Provisions for channel testing will be incorporated. The instrument power will be supplied from a vital bus. This will provide for a single indication in the control room of the auxiliary feedwater flow from each pump which is reactor protection system grade. The steam generator level indication (also reactor protection system grade) includes redundant channels and provides a diverse method of verifying flow. The operator also has pump discharge pressure (non-redundant and not safety grade) available to determine pump operation. This configuration will, therefora, provide the capability in the control room to ascertain the cct.,al performance of the auxiliary feedwater system. No additional mcdifications are required to satisfy the underlying NRC concerns in this area. Depending upon availability of supply, it is anticipated that the equipment will be operable by July 1, 1S80. 1211 098 2.1.7-2
2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.a Improved Post-Accident Samoling Capability A design and operational review of the reactor coolant and contain-ment atmosphere sampling systems will be performed at Point Beach Nuclear Plant to determine the ability of obtaining the desired samples under accident conditions without incurring radiation exposure to personnel in excess of applicable 10 CFR 20 limits. This review will entail the identification of involved systems, the derivation of an e'coropriate source term, and the performance of radiation shieldirr analyses as appropriate. The review will logically be performeu in conjunction with the shielding analysis for Item 2.1.6.b. 5hielded sampling and dilution stations will be provided if their need is demonstrated by the analysis. The review will include consideration of the ability to perform radiological and boron analyses. There does not appear to be any need for determining chloride during the course of an accident. It is anticipated that the review will commence befora January 1, 1980, that the review will be completed in mid-1980, and that any necessary modifications will be implmented by mid-1981. 1211 099 2.1.8-1
2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.b Increased Range of Radiation Monitors 1. An increased range gaseous effluent monitor capable of measuring as close to 105 uCi/cc ( Xe-13 3) as commercially available and feasible will be installed on the Auxiliary Building vent stack at Point Beach Nuclear Plant by January 1, 1981. The Gas Stripper Building vent stacks will be reviewed to determine the need for increased range monitoring. If needed, the equipment will be installed by January 1, 1981. 2. Radiciodine sampling on charcoal filters for plant gaseous effluents is presently available at Point Beach Nuclear Plant. The existing system will be reviewed to determine its adequacy under accident conditions. Modifications, if necessary, will be implemented by January 1, 1981. 3. Point Beach Nuclear Plant (PBNP) is not equipped with post-ace' dent, in-containment radiation monitoring devices. The Wisconsin Electric Power Company Three Mile Island Accident Review Task Force concluded that such instrumentation is neither appropriate, desired, nor needed, since there is no purpose for such information from the viewpoint of either public health and safety or operation and recovery. Contain-ment atmosphere radioactivity concentration data can be useful for both predicting offsite consequences and for providing in-plant radiological data, particularly when containment entry is desired. This information is obtainable by sampling the containment atmosphere, provisions for which already exist at PBNP; for example, t.be post-accident containment ventilation system is available as a sampling location. Direct radiation readings for post-accident containment entry are best obtained by the prudent and cautious use of portable instrumentation at the outer hatch, in the.. airlock, at the inner hatch, while cracking the inner hatchi while entering, and while inspecting. Remote instrumentation is not needed; if it were available, it would not be trusted, and primary information would still be obtained from portable instruments. The licensee, therefore, concludes that in-containment radiation level monitors are not needed at Point Beach Nuclear Plant. Further discussion of this subject is contained in the Wisconsin Electric Three Mile Island Accident Review Task Force Report at Section 5.3.8. 2.1.8-2 121i 100
2.1.9 ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS AND ACCIDENTS Transient and Accident Analysis The Westinghouse Owners' Group, of which Wisconsin Electric Power Company is a member, is performing generic analyses and evaluations of small br_ak loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and transients and accident scenarios which nelude operator actions not previously analyzed. The small break LOCA analyses have been completed and are reported in WCAP-9600, which was submitted to the Bulletins and Orders Task Force by the Owners' Group on June 29, 1979. Minor changes in LOCA emergency procedures resulting from the analyses and retraining of operators will be completed by January 1, 1980. The generic analysis of inadequate core cooling being performed by the Owners' Group will be completed in two phass. The first phase will be completed by October 31, 1979, and will consist of an analyses of a loss of primary system inventory resulting in core uncovecy with reduced heat removal capability from the core. Procedure guidelines using existing plant instrumentation will be developed to assist the operator in recognizing such an event and in determining what actions to take to assure core cooling. The Phase 1 analysis and guidelines are intended to be interim guide-lines and should not be considered final plant procedures. The guidelines will be implemented as Special Orders by January 1, 1980, and will act as interim instructions to the operator until the Phase 2 analyses and guidelines are completed. The second phase will be a more complete analysis of the conditions that could result in inadequate core cooling, and the development of procedural guidelines which will instruct the operator regarding the action needed to avoid or mitigate such a situation. It is expected that the Phase 2 analyses and procedural guideline implementation can be completed by the end of the first calendar quarter of 1980. The Owners' Group is performing a generic assessment of the adequacy of instructions to deal with transient and accident scenarios including operator actions not previously analyzed. This effort should be empleted in early 1980 with implementation of any procedural guidelines three months later. In addition to the above program, the Owners' Group is providing a pretest analysis, including a prediction of results, of the LOFT test program and semi-scale runs planned to be completed during November. The exact schedule for this item will be established with the Bulletins and Orders Task Force. 12ii 101 2.1.9-1
Containment Pressure, Water Level, and Hydrogen Monitors Wisconsin Electric Power Company will provide containment high range pressure, water level, and hydrogen monitors, if feasible, for the Point Beach Nuclear Plant. January 1, 1981 will be the target date for installation unless this cannot be met due to delays caused by long lead times for the delivery of equipment. Installation of this equipment will be accomplished during the earliest practicable refueling shutdown. A special plant shutdown solely for the purpose of installing this equipment in order to meet an arbitrary date is, we believe, inappropriate. Reactor Coolant System Venting The need for reactor coolant system venting is presently under review. This includes consideration of the vent location (pressurizer or vessel head), and how such a system would be operated if installed. The results of our review, as well as any proposed modifications to the plant, will be submitted to the NRC by January 1, 1980. The Wisconsin Electric Three Mile Island Accident Review Task Force discusses Reactor Coolant System Venting at Section 5.2.1. 1211 102 2.1.9-2
2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.c Improved In-Plant Iodine Instrumentation Existing equipment and procedures at Point Beach Nuclear Plant are adequate for the accurate determination of radiciodine under most conditions. However, procedures will be reviewed and revised as appropriate to emphasize the need for post-sampling ventilation of charcoal cartridges and gamma spectrum determination under accident conditions. An agreement between the Kewaunee Nuclear Plant and Point Beach Nuclear Plant will be completed to provide backup analytical laboratory services if required. These actions will be completed by January 1, 1980. l2 i 2.1.8-3 r
2.2.1
- PROVED REACTOR OPERATIONS COMMAND FUNCTION 2.2.1.a Shif t Supervisor's Responsibilities Point Beach Nuclear Plant (PBNP) operations are carried out in accordance with the Administrative Control Policies and Procedures Manual (QA volume 1), which specifies the responsibility and authority of each key position in the plant, including that of the Shift Supervisor.
The responsibility and authority of the Shift Supervisor are specified in Procedure PBNP 4.2 of QA volume 1, with the latest revision dated May 8, 1978. We will review this document for possible updating by January 1, 1980. A copy of the PBNP Administrative Control Policies and Procedures Manual is located at the NRC I&E offices in Glen Ellyn, Illinois, and updates are routinely provided directly to the NRC Compliance Inspector. The review of the duties, responsibilities, and authority of the Shift Supervisor and control room operators will emphasize those aspects addressed in Item 2 of the NRC position. The initial training and retraining programs for Shift Supervisors, as well as their administrative duties, will also be reviewed. These reviews and any required changes will be completed by January 1, 1980. 121i 104 2.2.1-1
2.2.1 IMPROVED REACTOR OPERATIONS COMMAND FUNCTION 2.2.1.b Shift Technical Advisor As stated in Appendix X of NUREG-0578, the intent of the position is to provide "that additional technical and analytical capability, dedicated to concern for the safety of the plant...in the control room to support the diagnosis of off-normal events and to advise the shift supervisor on actions to terminate or. mitigate the consequences of such events". It is further stated that, even if licensed, "In any event, when assigned as shift technical advisor, these personnel are to have no duties or responsibilities for manipulation of controls or command of operations". "The shift techical advisor would report to the shift supervisor in the control room during off-normal reactor plant conditions. It should be emphasized that the role of the shif t technical advisor is to serve in an advisory capacity to the shift supervisor and not to assume command or control functions. The shift supervisor may choose to direct the shift technical advisor to perform his advisory role from either the control room or the onsite technical support center, or the shif t supervisor may direct the shift technical advisor to serve as a liaison between technical support personnel manning the onsite technical support center and the shif t supervisor (see Section 2.2.2.b)". In order to best implement this recommendation and meet the above intent with properly qualified and trained (equal or equivalent per Section 2.2.1.b criteria) assigned personnel, Wisconsin Electric Power Company will implement a Duty and Call Technical Advisor position for the Point Beach Nuclear Plant. The primary responsibility of the Duty and Call, Technical Advisor will be to be available to the Shift Supervisor, when. called, to support the diagnosis of off-normal events, and to advise the Shift Supervisor of actions to terminate or mitigate the consequences of such events. Such events may include, but are not limited to, technical problems, reportable occurrences, or other significant operating events. Duty and Call Technical Advisor will be assigned to duty for seven ~ and will have available at least two continuous days and nights, systems of communication to the control room. Three communication systems will be provided: normal telephone, telephone " beeper" (seven-second message one way) system, and a portable radio (Motorola KRQ717) to the control room system. When "on duty", the Duty and Call Technical Advisor will be accessible for immediate communication from the plant at all times and will have the ability to be at the plant within 30 minutes after being called to return to the site in case of an emergency. Our Pcint Beach Nuclear Plant has had successful experience with the Duty and Call Superintendent roster in existence since the outset of licensed operations in 1970. Based on this experience, we consider i211 105 2.2.1-2
2.2.1.b (continued) the Duty and Call Techical Advisor response capability to report to the plant within 30 minutes both adequate and reasonable. At Point Beach Nuclear Plant, terrain is flat. Duty and Call Superintendents live within twelve miles of the plant, and auto transportation to the plant is very fast and unimpeded with multiple routes of travel of nearly equal distance. Over the nine years of Point Beach Nuclear Plant operation, there have been numerous occasions for the backshift Shift Supervisors to communicate with or call out a Duty and Call Superintendent. The system has worked very well, and there has been no occasion where the assigned Duty and Call Superintendent could not be reached, or did not respond to meet the need in the time frame required. We plan to continue our present Duty and Call Superintendent arrangement. The Duty and Call Technical Advisor will report to the Shift Supervisor in the control room during off-normal plant conditions. The Shift Supervisor or the Duty and Call Superintendent may choose to direct the Duty and Call Technical Advisor to perform his advisory role from either the control room or the On-Site Technical Support Center, or they may direct the Duty and Call Technical Advisor to serve as a liaison between technical support personnel manning the On-Site Technical Support Center and the control room. The Duty and Call Technical Advisor will report to his regularly assigned plant group head during normal plant operations when not on duty assignment. T ' Duty and Call Technical Advisor position requires a knowledge 'c fundamentals in the areas of reactor physics, chemistry, ta-. fluid mechanics, and heat transfer, in addition to the ap-- e' transient analysis and other safety analyses as applied to a Beach Nuclear Plant. A knowledge of Point Beach operating, and emergency procedures similar to that _o e a licensed reactor operator is also required. An v.u t . orf m or other scientific degree is desirable, but may be . en if s.ufficient nuclear-related experience has been obtained. assigned Duty and Call Superintendent and Duty and Call Technical isor will complement one another to provide total coverage of. . ant design, construction and operation for normal,~ transient:, dt "t - uccident conditions. The qualification basis for the Duty and Call Tcchnical Advisor is discussed further in the Attachment to 2.2.1.b. With the above arrangement of a 30-minute reporting commitment, Wisconsin 'lectric Power Company will assign Duty and Call Technical Advisor coverage from the presently employed Technical Assistant group at Point Beach Nuclear Plant by January 1, 1980. These personnel would meet the short-term qualifications of the Duty and. Call Technical Advisors, and it is projected that they could achieve full qualifications within one year (1/1/81). This position for a Duty and Call Technical Advisor is consistent with both the ACRS position expressed in the letter from M. Carbon (ACRS) to J. Hendrie (NRC), August 13, 1979, and its acccmpanying Enclosure 2 and the AI" Subcommittee on Operations position paper presented to the NRC Topical Meeting on October 12, 1979. 12ii 106 ~
- 2. 2.1-3
Attachment of 2.2.1.b Availability and Qualification Basis for Duty and Call Technical Advisor The accident or emergency response needed in the plant can be considered to consist of three response phases: 1. Immediate response necessitated by the indications or alarms, due to the initiating event or first effects of the off-normal condition. 2. Follow-up response and securement actions based on procedures and plant parameter trends. 3. Recovery response capability as determined by the plant status and systems availability. The expert capabilities needed are thus dependent upon the response phase. The physical presence of a Technical Advisor in the control room, on-shift, or on-site is not beneficial during the immediate response phase. The immediate response capability requires expertise in the direct operational chain of command to direct or manipulate plant controls. This portion of the NRC reccmmendation to improve reactor operation command functioning, where required, can be met only by improved training of licensed operators, improved procedures, and appropriate equipment. The presence of a Technical Advisor on shift in the control room during this initial phase of a transient may, in fact, detract from a clear definition of responsibility and authority, delay operator response, and contribute to increased control room confusion. The follow-up response and securement actions occur over an extended period of time and allow for utilization of other technical resources. An engineer or other technical assistance available on call can then supplement the operators and provide " additional technical and analytical capability, dedicated to concern for safety of the plant" with "no duties or responsibilities for manipulation of controls or command of operations". This provides the Shift Supervisor with a detached confirmation of the event diagnosis, an overview of the procedures used, including actions taken or in progress, and advice relative to termination or mitigation of the consequences of the event. The recovery response capability occurs over a much longer time period and may require diverse technical expertise to deal with details of the recovery process. Due to the longer ~ time period available for this response, the intent of the recommendation can be accomplished by utilizing staff and support resources in a planned mode for long-term accident mitigation and plant recovery operations. The technical qualifications for the Duty and Call Technical Advisor with a thirty-minute response capability can be met in the short term by upgrading the current plant Technical Assistants. The Technical Assistants are professional-level staff with B.S. or M.S. 1211 107 2.2<l-4
Attachment to 2.2.1.b (continued) degrees in engineering or the ociences and are assigned to work in specific areas of the plant. Several of the staff, because of their education and rotated assignments in the past at Point Beach, have already achieved a majority of all the requirements for a Technical 3dvisor. Further, continued work on the day shift will afford them the opportunity to be involved in the overall plant activities, meetings, training sessions, and development that only occur on the day shift. The need to keep abreast of the technology, in order to contribute as a Technical Advisor, ia recognized as an overriding reason to keep these personnel on the day shift, rather than semi-isolated on shif t work. In addition, shift work is usually unacceptable to professional personnel. The long-term complement of Duty and Call Technical Advisors is envisioned to consist of professional-level Nuclear Plant Engineers and as alternate members experienced supervisors who have had at least ten years of plant operation as licensed Senior Reactor Operators (SRO). Additional training of the experienced SRO-type personnel to meet the qualifications of the Technical Advisor position will be given as required on an individual basis. A college degree in engineering or science, therefore, may not be absolutely required for this position, but the fundamentals of an engineering degree needed to understand plant behavior during transients and accidents will be required. Additional training will stress these fundamentals as well as plant behavior during transients and accidents. The complement of Duty and Call Technical Advisors (DCTA) would perform the operating experience assessment function as part of their normal day shift duties. On a continuing basis, the DCTA's will be routed information and will be involved in the following areas of Point Beach operations and other reactor technical aspects through participation in and/or review of: a. Safety significant modification requests, b. Operating procedures, major maintenance procedures, I&C procedures, and emergency procedures and plans, c. Point Beach Nuclear Plant Licensee Event Reports and Significant Operating Events, and other applicable Licensee Event Reports, d. Point Beach Nuclear Plant period reports to federal and state regulatory bodies, NPRDS reports of component failures, e. f. Manager's Supervisory Staff Meeting Minutes, g. Safety-related maintenance requests, and h. Refueling and other major outage planning documents. 1211 108 2.2.1-5
2.2.1.c Shift and Relief Turnover Procedures Control room personnel (Shift Supervisor, Operating Supervisor, and Control Operators) utilize existing relief and turncver procedures which include review of control room logs, plant and equipment status, and work activities. These existing systems will be reviewed, in accordance with the intent of Item 1 of the NRC position, and checklists will be added where necessary by January 1, 1980. The present Auxiliary Operator logs are reviewed by the Auxiliary Operators coming on watch prior to shift relief. Additional signatures will be added to include the off-going watchstander. All major equipment out of service, both safety-related and non-safety-related, for a unit in operation is listed on the status board for that unit in the main control room. In all cases, the equipment moves in and out of service via the orderly command function of the Shift Supervisors. This responsibility does not reside in the Auxiliary Operator turnover and will, therefore, not be included. The existing surveillance sytem (which already includes period verification of the position of valves in all safety-related systems) will be modified to include any checklists to be developed, as well as any changes to the turnover procedures. This will provide for a period review and evaluation of these procedures. 9 i211 109 2.2.1-6
2.2.2 IMPROVED IN-PLANT EMERGENCY PROCEDURES AMD PREPARATIONS 2.2.2.a Control Room Access By special memorandum dated October 8, 1979, from the Manager, Nuclear Operations, to the Paint Beach Nuclear Plant Duty and Call Superintendents and the Duty Shift Supervisors, an admini-strative procedure has been implemented establishing the authority and responsibility of the person in charge of the control room to limit access during an abnormal operational transient or an accident to those individuals responsible for the direct operation of the Plant, to Technical Advisors, and to a predesignated NRC representative. The latter individual has not been designated at this time. This special memorandum on control room access policy will be incorporated into the Administrative control Policies and Procedures Manual (QA Volume 1) by January 1, 1980. Current lines of control room authority and line of succession are already included in the Point Beach Nuclear Plant Administrative Control Policies and Procedures Manual (QA Volume 1) and in the regulationa contained in 10'CFR 55. Those provisions have been reaffirmed in the memorandum referenced above. No further procedural modification is necessary to implement these . considerations. O 1211 110 2.2.2-1
2.2.2 IMPROVED IN-PLANT EMERGENCY PROCEDURES AND PREPARATIONS 2.2.2.b Onsite Technical Supeort Center Temporary Onsite Technical Support Center On a temporary basis, the existing administrative office area at Point Beach Nuclear Plant & 3.11 be designated as the Onsite Techical Support Center for use in the event of an accident. Although this location is not currently provided with emergency filtration or accident shielding, it has a number of advantages which make this location entirely appropriate. Current Plant drawings, telephones, restrooms, and kitchen facilities are already available. Access to the Plant switchboard is available. Portable electrical generators are available onsite to power small electrical appliances if required. Conference facilities are available. The location is a good staging area. The only additional modification required to activate this interim Onsite Technical Support Center is the relocation of one of the NRC. telephones to this office area and the provision of portable radiation monitoring instrumentation. These minor modifications will be completed by January 1, 1980. Appropriate modifications to the Point Beach Nuclear Plant Emergency Plan will be made. Permanent Onsite Technical Support Center A permanent Onsite Technical Support Center will be provided at Point Beach Nuclear Plant. Early planning for this facility has already begun and conceptual design will begin before the end of the year. The permanent Onsite Technical Support Center will have appropriate communications with the control room and the NRC, the capability to display selected Plant parameters, and emergency power supplies. The facility will be non-seismic. Nonredundant radiological filtration will be provided. It is anticipated that the detailed design will be completed in mid-1980 and that construction will be completed by January 1, 1982, subject to requisite regulatory approvals. Appropriate modifications to the Point Beach Nuclear Plant Emergency Plan will be made. 12il ii1 2.2.2-2
.2.2.2 IMPROVED IN-PLANT EMERGENCY >ROCEDURES.AND PREPARATIONS 2.2.2.c Onsite Operational Supoort Center Temporary Onsite Operational Support Center A temporary Onsite Operational Support Center will be established at Point Beach Nuclear Plant. This Center will serve primarily as a staging area for operations support personnel. This temporary Center will be located at the existir.g Energy Information Center (formerly the Information and Training Building). This facility is located onsite but outside of the fenced security area. It is equipped with restroom facilities, conference room facilities for up to 100 persons, kitchen facilities, outside telephones, and a Plant telephone, and an 'nlisted telephone. Long-term Plant records are stored in a vault in this building. The only modifi-cation required to activate this location as an interim Operational Support Center is the addition of a NRC telephone. This modifica-tion will be completed by January 1, 1980. Appropriate modifications to the Point Beach Nuclear Plant Emergency Plan will be made. Permanent Onsite Operational Support Center A permanent Onsite Operational Support Center will be designated at Point Beach Nuclear Plant. Planning for the permanent Onsite Operational Support Center is part of the p'lanning for the Onsite Technical Support Center. The final designated location may or may not be part of the Onsite Technical Support Center. As indicated in the response to 2.2.2.b, overall planning is expected to be completed and conceptual design is expected to commence by January 1, 1980. Final designation of the Onsite Operational Support Center may not take place before completion of the Onsite Technical Support Center, anticipated by January 1, 1982. Appropriate modifications to the Point Beach Nuclear Plant Emergency Plan will be made. 1211 1I2 2.2.2-3
,e RESPONSES TO ENCLOSURE 7 NEAR TERM REQUIREMENTS FOR IMPROVING EMERGENCY PREPAREDNESS 1. The Point Beach Nuclear Plant Emergency Plan was reviewed by the NRC Division of Reactor Licensing in 1975. By letter dated October 24, 1975, the NRC concluded "that the Plan meets the requirements of 10 CFR 50, Appendix E, and therefore provides an adequate basis for an acceptable state of emer- .gency preparedness". We, therefore, conclude that no further modifications to the Emergency Plan are appropriate at this time, except for those needed to address the implementation of Operational and Technical Support Centers. 2. Implementation of certain short term actions recommended by the Lessons Learned Task Force has already been addressed in our response to Item 2.1.8. 3. Planning and location selection for an Emergency Operations Center for federal, state, and local personnel will be done concurrently with the planning for the Onsite Technical Support Center and the Operational Support Center as described in our responses to 2.2.2.b and 2.2.2.c. Accordingly, the establishment of a permanent Emergency Operations Center will take place by January 1, 1982, as noted. 4. A review of existing off-site monitorin'g capabilities will be performed early in 1980. Additional monitoring equipment, if required, will be provided by mid-1980. 5. No action required by Licensee. 6. No action required by Licensee. 0 4 7 E7-1}}