ML19250A759
| ML19250A759 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck, Millstone File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/18/1979 |
| From: | Counsil W NORTHEAST UTILITIES |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-42588, NUDOCS 7910240515 | |
| Download: ML19250A759 (49) | |
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J October 18, 1979 Docket Nos. 50-213 50-245 50-336 Darrell G. Eisenhut, Acting Director Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Reference:
(1)
D. G. Eisenhut letter to All Operating Nuclear Power Plants dated September 13, 1979.
Gentlemen:
Haddam Neck Plant Millstone Nuclear Power Station, Unit Nos.1 and 2 Followup Actions Resulting From the NRC Staff Reviews Regarding the Three Mile Island, Unit 2 Accident In Reference (1), Connecticut Yankee Atomic Power Company (CYAPCO) and Northeast Nuclear Energy Company (NNECO) were requested to respond by providing commitments to comply with each of the requirements of Enclosure 6 of that Reference.
In response to that request, CYAPCO hereby provides Attachment 1; and NNECO hereby provides Attachments 2 and 3 for Millstone Unit No. 1 and Millstone Unit No. 2, respectively, in response to each of the requirements of NUREG-0578 as amended by Reference (1).
The Attachments are self-contained with respect to CYAPCO's and NNECO's positions on each of the requirements; however, the following general comments are provided to clarify the origin of these positions.
It is appropriate to state that CYAPC0 and NNECO are in conceptual agreement with the safety concerns identified in NUREG-0578 without exception. However, there are instances where there is not agreement as to how these concerns should be addressed; and in these cases, ample justification for the proposed alternative is provided. Thus, in all cases, CYAPC0 and NNECO have proposed the optimum method of addressing the concern, in recognition of plant-unique features, concurrent related engineering evaluations, and supplemental clarifying information which has been obtained to date.
It is also accurate to characterize CYAPCO's and NNECO's ef forts to implement the requirements in accordance with the schedule of Enclosure 6 of Reference (1) as unparalleled by any previous response to regulatory requirements. However, current estLnates indicate that especially for certain hardware-related items, circumstances beyond the control of CYAPCO and NNECO may preclude total implemen-tation by January 1,1980 for the Category A items. This date is also undul*;
SIT 7910240 sta 1205 120 /
premature where total implementation necessitates a plant shutdown. As indi-cated in CYAPCO's and NNECO's monthly operating reports, the refueling outages for the three operating units are currently scheduled for mid-1980. CYAPCO and NNECO have determined that mid-cycle shutdowns exclusively to implement Reference (1) are inappropriate. The economic impact of January 1, 1980 shut-downs, especially in the New England area with its relatively high percentage of nuclear generating capacity, is not of f set by the advantages of earlier implemen-tation of the modifications outlined in the Attachments.
NUREG-0578 items which fall into the above set of circumstances by current estimates are indicated in the attached material.
In summary, CYAPC0 and NNECO are committed to implement the modifications as expeditiously as practical. Nonetheless, justifiable alterna-tives to mid-cycle shutdowns will be pursued where circumstances so dictate.
Should currently unscheduled plant outages occur before the refuelings, modifica-tions will be completed to the extent possible during those periods consistent with design and hardware availability. As an example of how this concept will be applied, efforts are in progress to Laplement the recommendations of NUREG-0578, Item 2.1.3, at Millstone Unit No. 2 during the upcoming outage scheduled to begin October 27, 1979.
In addition, CYAPC0 and NNECO will make every ef fort to comply with the requirements of Enclosure 7 to Reference (1) in accordance with the schedule of Enclosure 8.
However, the degree to which this will be fulfilled is dependent upon the timeliness of the NRC guidance on specific criteria and its adaptability to site-specific feature.
This concern is especially applicable to Items 1 and 3(b) on Enclosure 8.
In the case of Items 5(b), 6(b), and 6(c), CYAPCO and NNECO will ensure that the revised NRC intent and guidance is conveyed to appropriate State and local agencies / officials. The State of Connecticut and the towns have emergency plans with which the NRC has concurred. CYAPCO and NNECO will continue to cooperate with the State and towns in their efforts to evaluate and implement the require-ments in the area of emergency preparedness.
CYAPCO's and NNECO's continuing dialogue with the NRC Staf f via Owners Group participation, topical meetings, and other sources will undoubtedly supplement the Attached material as the implementation dates approach.
In response to the Reference (1) request, it is emphasized that CYAPCO and NNECO are indeed committed to resolve each of the Reference (1) items to our mutual satisf action, on a schedule which recognizes the realistic constraints of hardware availability, manpower limitations, and system reserve generating capacity, as well as the safety signi-ficance of these items.
We trust you find the above information responsive to your request.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY NORTHEAST NF EEAR ENERGY COMPANZ 7
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- 4. G. Counsil Vice President 1205 121
DOCFIT No. 50-213 ATTACHMENT 1 HADDAM NECK PLAhT FOLLOWUP ACTIONS RESULTING FROM THE NRC STAFF REVIEWS REGARDING THE THREE MILE ISLAND, UNIT 2 ACCIDENT 1205 122 OCTOBER, 1979
i HADDAM NECK PLANT FOLLOWUP ACTIONS RESULTING FROM THE NRC STAFF REVIEWS REGARDING THE THREE MILE ISLAND, UNIT 2 ACCIDEhT Item 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief and Block Valves, and Pressurizer Level Indicators in PWR's.
CYAPCO Response The redundant emergency power requirements for pressurizer heaters, power-operated relief valves and a sociated block valves, and pressurizer J evel indication instrument channels will be met on Haddam Neck Plant in conformance with requirements appropriate for this unit. These functional requirements will be submitted for the Lmple=entation Review no later than January 1,1980.
CYAPCO intends to implement these requirements by January 1,1980.
Item 2.1. 2 Performance Testing for BWR and PWR Relief and Saf ety Valves.
CYAPCO Response A program for testing power operated relief valves (PORV's) and saf ety valves (SV's) used for primary system pressure control under design bases operating conditions is being developed by the NSSS Owners Group. This program includes definition of test conditions and qualification requirements for all specified valves in operating reactors. The results of this program will be made available to the generic efforts being undertaken by the industry (through for example the Electric Power Research Institute, EPRI, and the Nucles: lafety Analysis Center, NSAC) no later than January 1,1980. These results will a!,a be available for discussions with the NRC Staf f to establish Generic Resolutions no later than January 1, 1980. The Haddam Neck Plant will comply with the schedule for completion of the test program which is agreed to during these Generic Resolution meetings.
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_2 Item 2.1.3 Information to Aid Operators in Accident Diagnosis and Control.
a) Direct Indication of Power-Operated Relief Valve and Saf ety Valve Position for PWR's and BWR's.
CYAPCO Response The Haddam Neck Plant primary system relief and safety valves will be provided with a position indication in the control room derived from either a reliable valve position indication device or a reliable indication of flow in the discharge pipe. The functional requirements and conceptual design will be submitted for the implementation Review no later than January 1,1980.
CYAPC0 intend; to implement this requirement and submit the design details to the NRC by January 1,1980, to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.3 Information to Aid Operators in Accident Diagnosis and Control.
b) Instrumentation for Detection of Inadequate Core Cooling for PWR's and BWR's.
CYAPC0 Response The procedures to be used by an operator to recognize inadequate core cooling will be developed based on analyses being performed as required by Item 2.1.9, Transient and Accident Analysis, Analysis of Inadequate Core Cooling; these analyses are described below. The guidelines for the procedures are being developed by the NSSS Owners Group and will be available for discussions with the NRC Staf f to establish Generic Resolutions no later than January 1,1980.
If the analyses or the guidelines indicate the need for the design of new instrumentation, the design of su-h instrumentation will be made available for discussions with the NRC Staf f to establish Generic Resolutions.
The functional requirements and a conceptual design for a reactor vessel level measurenant device are being developed by the NSSS Owners Gro This effort 120E 124
. includes e survey of currently available technology and assessment of the f easibility of various alternatives.
If required by the Generic Resolution discussions with the NRC Staf f, the functional requirements and conceptual design will be submitted for Proposal Review by the NRC Staf f prior to implementation.
The installation schedule for such a device, should it be deemed necessary, will be established during the Proposal Review.
CYAPC0 will install a safety grade primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that it is not to be used exclusive of other related plant parameters.
CYAPCO intends to implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.4 Containment Isolation Provisions for PWR's and BWR's.
CYAPCO Response The Haddam Neck Plant containment isolation system vill comply with the recommendations of SRP 6.2.4 regarding diversity in the parameters sensed for the initiation of containment isolation. A generic review of all systems pene-trating the containment on all operating plants is being conducted by the NSSS Owners Group. This review will produce generic criteria for the definition of essential systems, identification of all such systems, and specification of the bases for each system's selection. Criteria are also being developed for selective unisolation of non-essential systems which may be beneficial. The results of this review will be submitted for Implementation Review no later than January 1, 1980.
It is the intention of CYAPC0 to provide diverse signals for containment isolation by January 1,1980 to the extent possible in consideration of the scheduling of constraints discussed in the forwarding letter, and to review the adequacy of the existing containment isolation system in accordance with the guidance in NUREG-0578. However, the current broad scope of SEP Topic VI-4, Containment Isolation, is such that any other potential modifications should be deferred until SEP topic resolution is achieved. The SEP Topic includes such 1205 125
aspects as conformance to 10CFR50, Appendix A, General Design Criteria 54 through 57, ability of the purging / ventilation syster. Isolation valves to close upon receipt of an accident signal, use -f rest.lient sealing material s, the adequacy of the maintenance and repair sthedules, etc.
The current NRC Staff echedule of evaluating this topic is such that significant delays from the NUREG-0578 schedule are not anticipat ed.
The efficiency of integrating SEP and TMI-related evaluations and resulting modifications warrants deferral in this case.
It em 2.1. 5 Post-Accident Hydrogen Control Systems for PWR and BWR Containments.
a) Dedicated Penetrations for External Recombiner or Post-Accident External Purge System b)
Inerting BWR Containments c) Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant CYAPCO Response Haddam Neck Plant currently meets the intent of the Item 2.1.5a requirement with a manual locked closed valve.
It is also noted that the NRC has delayed i
proposed rulemaking proceedings on Items 2.1.5b and 2.1.5c, requiring no action on these items at this time.
Item 2.1.6 Post-Accident Control of Radiation in Systems Outside Containment of PWR's and BWR's.
a)
Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems)
CYAPC0 Response CYAPC0 will develop and Laplement a program to reduce leakage from systems outside containment that would or could contain high-level radioactive materials during a serious transient or accident to as low as practical levelc. The program will include initial implementation of all practical leak reduction measures on the af fected systems and measurement of actual leakage rates with the systems in operation. A continuing program will also be 1205 126'
I.
established to implement preventative maintenance to reduce leakage and periodic leak testing Et a frequency not to exceed refueling outage intervals.
The systems to which this program applies will be determined based upon the syste= function during and after a serious transient or accident.
CYAPC3 intends to bnplement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Iten 2.1.6 Posc-Accident Control of Radiatioa in Systems Outside Containment of PWR's and BWR's, b) Design Review of Plant Shielding of Spaces for Post-Accident Operations.
CYAPCO Response CYAPCO will perform a radiation and shielding design review of the spaces around systems which may contain highly radioactive materials as the result of an accident. The review will identify the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipnent may be unduly degraded by the radiation fields during post-accident operations cf these systems. The design review shall determine the corrective actions needed for vital areas throughout the plant.
Radiation level guidance and shielding source term criteria for this design review will be in accordance with the information presented by the NRC at the Region 1 Meeting on September 24, 1979. CYAPCO intends to Lnplement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Should these reviews identify any significant deficiencies, such as high radiation in the control room, modifications will be performed on an expedited basis.
However, lesser deficiencies will not be resolved until an integrated assessment of overlapping SEP topics is completed.
This def erral is justified by the potential for major structural, mechanical, and electrical modifications resulting from a variety of SEP Topics including:
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I 4
SEP TOPIC TITLE (1)
III-4.A Tornado Missiles (2)
III-4. B Turbine Missiles (3)
III-4.C Internally Generated Missiles (4)
III-4.D Site Proximity Missiles (5)
III-5.A Fipe Break Inside Containment (6)
III-5.B Pipe Break Outside Containment (7) III-6 Seismic Design Considerations (8) 111-12 Environmental Qualification (9)
VI-2.B Subcompartment Analysis (10) VI-8 Control Room Habitability Iten 2.1.7 Improved Auxiliary Feedwater System Reliability for PWR's a) Automatic Initiation of the Auxiliary Feedwater System CYAPCO Response Automatic initiation of auxiliary feedwater is considered to be unnecessary for the Haddam Neck Plant for the following recsons:
(1) immediate actions are required by the reactor trip procedure to verif y feedwater flow status; (2) there is complete control of the auxiliary f eedwater system from the main control board; (3) there is approximately forty (40) minutes available before auxiliary feedwater is required; and (4) past experience with tecovery from feedwater system problems indicates no need for automation of the auxiliary feedwater system.
Each of these facts was presented in detail to the Staf f in the responses to the TMI-related I&E Bulletins and follow-up correspondence.
This resulted in the current operating procedures which require one of the two control room operators to immediately initiate auxiliary feedwater if required.
It is not credible to assume that auxiliary feedwater will not be initiated within 40 minutes if it is required.
It is also noted that several other aspects of the issue of automatic initiation - f auxiliary feedwater should be considereu in the Staf f evaluation. Automatic initiation of auxiliary feedwater has the potential to complicate the feedwater line break event by increasing the cooldown rate of the primary system and increasing the severity of the environment within containment. The time required to address the above concerns renders the commitment to install automatic initiation particularly premature.
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Ongoing SEP topic evaluations also af fect CYAPCO's evaluation of this recommenda-The safe shutdown and high-energy pipe break reviews are based in part on tion.
the design of the auxiliary feedwater system. A larger impact is evidenced in the area of fire protection-related saf e shutdown modifications. CYAPCO has previously Should the provided the Staff with a schedule for an alternate shutdown system.
integrated avaluation of these topics result in the inntallation of an electrically driven auxiliary feedwater pump (s), the question of the adequacy of onsite emergency power supplies must be considered. Thus, SEP-related issues further support CYAPCO's position that it is premature to commit to this NUREG recommendation.
Adequate assurance is provided that auxiliary feedwater will be supplied to the Steam Generators w' en required.
Item 2.1.7 Improved Auxiliary Feedwater System Reliability for PWR's.
b) Auxiliary Feedwater Flow Indication to Steam Generators CYAPCO Response The Haddam Feck Plant auxiliary feedwater system will be provided with a control The grade indication of flew to each steam generator in the control room.
functional requirements for such an indication including the number of channels, range, control functions, alarm functions, display, recorded outputs, and setpoints are being developed. These functional requirements will be submitted for implementation Review by the NRC Staff no later than January 1, 1980.
CYAPCO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints noted in our forwarding letter.
The Haddam :Wck Plant auxiliary feedwater system will be provided with a safety grade indication of flow to each steam generator in the control room.
These The functional requirements for such an indication are being developed.
functional requirements will be submitted for Implementation Review by the NRC Staff no later than January 1, 1981.
CYAPCO intends to implement this require-ment by January 1,1981 to the extent possible in consideration of the scheduling constraints noted in our forwarding letter.
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Item 2.1.8 Instrue ntation to Follow the Course of an Accident a) Improved Post-Accident Sampling Capability CYAPCO Response CYAPCO will perform design and operational reviews of the reactor coolant and containment atmosphere sampling systems, radiological spectrum analysis f acilities, and chemical analyses capabilities in accordance with the require-ments of NUREG-0578. CYAPC0 intends to implement these ruquirements by January 1,1980, to the extent possible in consideration of the scheduling constraints noted in the forwarding letter.
In accordance with the philosophy presented in the response to Item 2.1.6.b, procedural improvements or modifications which are judged to be independent of SEP topics will be implemented without consideration to the SEP schedule.
However, potential major modifications which may be required will be deferred until an integrated assessment of the related SEP topics, identified in 2.1.6.b, is completed.
The implementation schedule for any required design modifications will be addressed in the January 1, 1980 submittal.
Item 2.1.8 Instrumentation to Follow the Course of an Accident b) Increased Range of Radiation Monitors CYAPCO Response CYAPCO will implement the requirements of NRC Position 2.1.8b, Items 1, 2, and 3 as modified in the NRC Region 1 Meeting on September 24, 1979 for high-range noble gas effluent monitors, high-range contain=ent radiation monitors, and high-range ef fluent radioiodine and particulate sampling and analysis. CYAPCO intends to Laplement this requirement by January 1,1981 to the extent possible consistent with commercial availability of equipment and the scheduling constraints discussed in the forwarding letter.
1205 130 In addition, CYAPCO will develop and implement procedures for estimating noble gas and radioiodine release rates if the existing effluent instrumentation goes off-scale.
CYAPCO intends to Lnplement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints
discussed in our forwarding letter.
Item 2.1.8 Instrumentation to Follow the Course of an Acciden; c)
Improved in-Plant Iodine Instrumentation CYAPCO Response CYAPCO will provide equipment, training, and procedures for determining air-borne iodine concentration throughout the plant under accident conditions in accordance with NUREG-0578 requirements. CYAPC0 wfll implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.9 Analysis of Design and Off-Normal Transients and Accidents CYAPCO Response The response to Transient and Accident Analysis requirements is being developed by the NSSS Owners Group in conjunction with General Resolution meetings with the NRC Bulletins and Orders Task Force. These responses will be submitted on the schedule agreed to by that Task Force and the NSSS Owners Group and will be referenced for specific application. The implementation of emergency procedures and retraining will be done by CYAPC0 on a schedule consistent with that established for the analysis requirements.
Additional Instrumentation, Containment Pressure, Containment Water Level and Hydrogen Monitors, to follow the Course of the Accident CYAPCO Response CYAPC0 will implement this requirement in accordance with the criteria discussed extent possible in consideration of the scheduling constraints discussed in our forwarding letter.
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Installation of Remotely Operated High Point Vent in Reactor Coolant System CYAPCO Response CYAPC0 will implement this requirement in accordance with the criteria discussed at the October 11, 1979 NRC Topical Meeting on this subject. CYAPCO intends to implement this requirement by January 1,1981 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Ite= 2.2.1 Improved Reactor Operations Command Function a) Shift Supervisor Responsibilities CYAPCO Response CYAPCO will implement the intent of this requirement.
However, in order to clarify the meaning of the term " accident situation" in Item 2.b of the Staf f 's position in Appendix A of NUREG-0578, the requirement is interpreted as follows: The Shift Supervisor or Supervising Control Operator shall until properly relieved remain in the control room at all times whenever a site or general emergency has been declared, to direct the activities of control room operators.
CYAPC0 intends to implement this requirement by January 1, 1980.
Item 2.2.1 Improved Reactor Operations Command Function b)
Shif t Technical Advisor Discussion Lmplementation of the Shif t Technical Advisor (STA) as proposed by the Lessons Learned Task Force would place a graduate engineer independent and detached from plant operations in the control room at or shortly following the occurrence of an accident or abnormal transient.
Because the STA would not be in the direct operational chain of command and, in f act, would not need to be licensed, he could neither manipulate nor direct licensed operators to manipulate the controls of the reactor plant.
He would be empowered to advise operations but not responsible to operations for his advice.
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The Shif t Supervisor is correctly charged with the responsibility for safe operation of the plant at all times. During the early phase of an accident, he discharges this responsibility by coordinating and directing the response of the control room staff. The actions of the operators are procedural, being governed by their training and emergency procedures, and during this phase, the entire control room staff including the shift supervisor is completely occupied Plant operating experience indicates that with responding to the accident.
there is a period of time following initiation of any accident or transient wherein the Shif t Supervisor has suf ficient time to analyze, diagnose, and respond to the condition of the plant but does not have sufficient time to carefully consider an independent assessment of the accident, resolve any conflicts between his and decide to alter the independent assessment and, on the basis of such assessment, the procedural actions of the operators. Dialogue regarding such an assessment or time spent resolving such conflicts can only distract and delay the Shif t Supervisor and consequently degrade the response of the control room staf f to the accident.
Even though the roles of Shift Supervisor and STA can be carefully delineated by procedure and training, industrial and military experience indicates that a direct-line organization wherein authority and responsibility are inter-The dependent is required to effectively operate in a crisis environment.
proposed STA is empowered to advise operations but not responsible to operations for his advice.
His authority and responsibility are not inter-dependent. A potential for conflict and confusion exists which cannot be completely eliminated by procedure or training because procedure and training can address only those event sequences which have been postulated in advance.
One important lesson learned from the experience at Three Mile Island and at other facilities is that not all event sequences can be postulated in advance.
Therefore, an alternative which avoids this potential for conflict and conf usion but improves the functions intended by the proposed STA is recommended.
Two functions are intended to be improved by the proposed STA:
(1) accident assessment and (2) operating experience assessment.
In order to improve the accident assessment function while avoiding the degradation in accident response which accompanies the proposed STA, the cocrse of an accident is considered in three sequential phases:
immediate, intermediate, and recovery.
The immediate phase extends from the point at which an abnormal condition affecting plant safety can be detected in the control room until the point at which the Shift Supervisor has suf ficient time to carefully consider an independent assessment and, on the b sis of such assessment, decide to alter the procedural actions of the operators. The intermediate phase extends from the end of the kmnediate phase until the point at which the Technical Support Center (TSC) is manned and ready. The recovery phase extends from the end of the intermediate phase until the point at which recovery is complete.
For the tcmediate phase, the accident assessment function can be improved only by upgraded training to enhance the operators' abilities to recognize, diagnose, and respond to accident conditions. During this phase, the operators' actions are governed by training and emergency procedures, and by definition, there is insuf ficient time for the careful consideration of an independent assessment which would be required before such an assessment could become the basis for altering the procedural actions of the operators.
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For the intermediate phase, the accident assessment function can be improved by either of two alternative means. An operator can be educated in science and engineering in order that he might provide an assessment which could be considered and acted upon by the Shift Supervisor. Alternatively, a graduate engineer or equivalent can be trained in plant operations and r de available to the Shift Supervisor on call in order that he might provide suc, an assessment.
In either case, the Shift Supervisor must have sufficient time to carefully consider the assessment and, based on such assessment, decide to alter the procedural actions of the operators.
For the recovery phase, the accident assesscent function can be improved by manning the TSC.
The collective engineering resource within the TSC will be able to develop a detailed independent assessment of plant conditions and provide appropriate procedures with which to recover from the accident.
The operating experience assessment function can best be provided by a team which reviews operating experience at the plant and at plants of like design.
Varying team membership as appropriate to the operating experience being assessed assures accomplishment of this function by the best qualified indi-viduals.
CYAPCO Response The two functions intended to be improved by the proposed STA will be implemented as follows:
(1) Accident Assessment a.
Immediate Phase
?
An operator or supervisor in the direct operational chain of command on each shif t (normally in charge in the control room) vill receive additional specific training in the response and analysis of the plant for transients and accidents.
This training vill be coordinated with the schedule for preparation and review of analysis and guidelines under the NRC Bulletins and Orders Task Force.
All operators and supervisors will receive 9dditional training appropriate to their responsibilities in the response of the plant to transients and accidents. This longer term training and qualification criteria will be provided by the Institute of Nuclear Power Operations.
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. b.
Intermediate Phase CYAPCO intends to implement the intermediate phase requirement for additional accident assessment capability by adding an addi-tional licensed operator to each shif t and upgrading the training of a senior reactor operator on each shif t to include the general technical education, and additional transient and accident response training requirements as discussed in Enclosure (2) to D. G. Eisenhut's letter of September 13, 1979. The senior reactor operator on each shift designated as the Shif t Technical Advisor would have no assigned line functions while performing the Shif t Technical Advisor function.
The addition of a licensed reactor cperator to the control room shift complement will enable the STA to be detached from controls manipulation and supervision of operators during an event. This provision is considered critical to our intended method of implementation since the primary deficiency noted in the Staff's discussion regarding this alternative was the need for involvement of each of the current shift complement of three operators in satisfying the demands for prompt control and supervisory actions.
Implementation of the STA requirement in this way also prevents dilution of command authority during an accident situation which was noted as not desirable in Enclosure (2) to D. G. Eisenhut's letter of September 13, 1979.
CYAPC0 feels that this method represents the optimum alternative for implementation of the STA requirement. However, as discussed in Enclosure (2) referenced above, the completion of the additional general technical education and transient and accident training requirements may take two years or more to fully tmplement depending on the scope and content of the training requirements as finally established. Therefore, in the interim period while the designated operators are off-shift or off-site receiving the required training, CYAPCO intends to implement an interim method of providing the additional accident assessment capability.
The interim method which CYAPCO intends to bnplement by January l',
1980 is to provide immediate on-call assistance to the control 1205 135
room by designated senior personnel from the plant staff available on-site in approximately 30 minutes. The majority of designated individuals will have a Bachelor's Degree in science or engineering and all will have a current Senior Reactor Operator License on the designated unit. These individuals will also rece'ive short-term supplemental training and retraining in plant transient and accident As an additional interim measure, CYAPC0 will also provide response.
an on-call group of experts with experience and technical backgrounds in the various technical areas important to safety including mechanical, electrical, and fluid systems; reactor physica, chemistry, and metallurgy. This group would consist of individuals from both the plant staf f and our Northeast Utilities Service Company (NUSCO) engineering staff.
Due to the close geographical proximity of the NUSCO engineering support group, this on-call team of experts will be available on-site within approximately one hour of an ever.t.
c.
Recovery Phase Individuals knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident will be available on-call to staf f the On-Site Technical Support Center.
(2) Operating Experience Assessment A team will be designated by CYAPCO/NUSCO to assess the operating experience at the Haddam Neck Plant and at plants of like design. Team membership may vary as appropriate to the operating experience being assessed but will include degreed engineers with experience in the various technical areas important to safety including mechanical, electrical and fluid systems engineering, reactor physics, chemistry, and metallurgy.
In addition, the team membership will include or provide routine access to persons experienced in the principles of human engineering or human factors. Procedures will be provided to insure close coupling of the results of the operating experience assessment function with the SIA function. The operating experience assessment function will be'imple-mented by January 1, 1980.
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. Item 2.2.1 Improved Reactor Operations Command Function c) Shift and Relief Turnover Procedures CYAPCO Response CYAPCO will review and revise plant procedures as necessary to comply with this requirement. CYAPCO intends to implement this requirement by January 1, 1980.
Item 2.2.2 Improved In-Plant Emergency Procedures and Preparations a) Control Room Access CYAPCO Response CYAPCO will review and revise plant emergency procedures as necessary to comply with this requirement.
CYAPCO intends to implement this requirement by January 1, 1980.
Item 2.2.2 improved In-Plant Emergency Procedures and Preparations b) Onsite Technical Support Center CYAPC0 Response CYAPCO will implement the requirement for an Interim Technical Support Center, in accordance with revised NUREG requirements, as follows:
(1) A location will be designated in the emergency plan. This may be a temporary location.
(2) Telephone communications will be established to the control raom and the NRC.
These may be temporary.
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. (3) The staffing and activation criteria will be specified in the emergency plan.
(4) The TSC will have access to plant technical information.
CYAPC0 intends to implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discus. sed in the for-warding letter.
CYAPCO will submit a preliminary design package for the final TSC by January 1, 1980.
Ite= 2.2.2 Improved In-Plant Emergency Procedures and Preparations c) Onsite Operational Support Center CYAPCO Response CYAPCO intends to comply with the requirements of this recommendation by January 1, 1980.
Item 2.2.3 Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability CYAPCO Response implementation of this requirement has been deferred by the NRC.
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e DOCl:ET No. 50-245 ATTACIMENT 2 MILLSTONE NUCLEAR POWER STATION, UNIT NO.1 FOLLOWUP ACTIONS RESULTING FROM THE NRC STAFF REVIEWS REGARLING THE THREE MILE ISLAND, UNIT 2 ACCIDENT 1205 139 OCTOBER, 1979
Millstone Unit 1 Responses to NUREG 0578, Lessons Learned Short Term Recommendations Reference 1:
BWR Owners group letter October 17, 1979 - T. D. Keenan to D. G. Eisenhut 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief and Block Valves, and Pressurizer Level Indicators in PWRs.
NNECO Response NNECO concurs with owners group response (Ref. 1) 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves.
NNEC0 Response NNEr0 concurs with the owners Group Response (Ref. 1) except as follows:
BWR Owner's Group Implementation Criteria:
i Item (3):
" Control grade systems, actuated by reactar vessel high water level, shall be provided to prevent the feedwater and high pressure injection systems from overfilling the vessel."
NNECO's Position NNECO is evaluating the advisability and methodology for backfitting an automatic system that would terminate feedwater and FWCI flow.
Our concern is that such a system will increase the likelihood of an inadvertent loss of feedwater and high pressure ECCS.
Therefore, we cannot commit to provide such a system without considerable further evaluation of design requirements and total integrated safety objectives.
2.1.3.A
" Direct Indication of Power-0perated Relief Valve and Safety Valve Position for PWR's and BWR's.
1205 140 NNECO Position NNECO concurs with owners group response (Ref. 1) and intends to install the sensing devices, by January 1, 1980, to the extent possible in consideration of the scheduling constraints discussed in cur forwarding letter.
2.1.3.B
" Instrumentation for Detection of Inadequate Core Cooling in PWR's and BWR's".
NNECO Response NNECO concurs with the owners group position (Ref.'1) 2.1.4
" Containment Isolation Provisions for PWR's and BWR's" NNECO Response NNECO concurs with the owners group position (Ref. 1) as follows:
It is the intention of NNECO to provide diverse signals for containment isolation by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in our forwarding letter, and to review the adequacy of the existing containment isolation system in accordance with the guidance in NUREG-0578.
However, the current broad scope of SEP Topic VI-4, Containment Isolation, is such that any other potential modifications should be deferred until SEP topic resolution is achieved.
The SEP Topic includes such aspects as conformance to 10CFR50, Appendix A, General Design Criteria 54 through 57, ability of the purging / ventilation system isolation valves to close upon receipt of an accident signal, use of resilient sealing materials, the adequacy of the maintenance and repair schedules, etc.
The current NRC Staff schedule of evaluati'g this topic is such that significant delays from the NUREG-0578 schedule are not anticipated. The efficiency of integrating SEP anu TMI-related evaluations and resulting modifications warrants deferral in this case.
2.1.5.A Post-Accident Hydrogen Control Systems for PWR and BWR Containments.
Dedicated Penetrations for External Recombiner or Post-Accident -
External Purge System NNECO Resoonse The actions required for this item will be integrated within the Systematic Evaluation Program (SEP) under the topic " Combustible Gas Control".
2.1.5.b & c Post-Accident Hydrogen Control Systems for PWR and BWR Containments.
b.
Inerting BWR Containments Capability to Install Hydrogen Recombiner at Each Light Water c.
Nuclear Power Plant 1205 141
NNEC0 Response These items have been deferred for evaluation by the NRC.
2.1.6.a Post-Accident Control Radiation in Systems Outside Cont,ainment of PWRs and BWRs.
Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems)
NNECO Response NNECO will develop and implement a program to reduce leakage from systems outside containment that could contain high level radioactive materials during a serious transient or accident to as low as practical levels.
The program will include initial implementation of all practical leak reduction measures on the affected systems and measuren,ent of actual leakage rates with the systems in operation.
A continuing program will also be established to implement preventative maintenance to reduce leakage and periodic leak testing of a frequency not to exceed refueling outage intervals.
The systems to which this program applies will be determined based upon the system function during and after a serious transient or accident.
NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
2.1.6.b Post-Accident Control of Radiation in Systems Outside Containment of PWRs and BWRs b.
Design Review of Plant Shielding of Spaces for Post-Accident Operations NNECO Response NNECO will perform a radiation and shielding design review cf the spaces around systems which may contain highly-radioactive materials as the result of an accident.
The review will identify the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.
The design review shall determine the corrective actions needed for vital areas throughout the plant.
Radiation level guidance and shielding source term criteria for this design review will be in accordance with the information presented by the NRC at the Region 1 Meeting on September 24, 1979.
NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
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Should these reviews identify any significant deficiencies, such as high radiation in the control room, modifications will be performed on an expedited basis.
However, lesser deficiencies will not be resolved until an integrated assessment of overlapping SEP topics is completed.
This deferral is justified by the potential for major structural, mechanical, and electrical modifications resulting from a variety of SEP topics including:
SEP TOPIC TITLE (1) III-4.A Tornado Missiles (2)
III-4.8 Turbine Missiles (3) III-4.C Internally Generated Missiles (4) III-4.0 Site Proximity Missiles (5) III-5.A Pipe Break Inside Containment (6)
III-5.B Pipe Break Outside Containment (7) III-6 Seismic Design Considerations (8) III-12 Environmental Qualification (9) VI-2.B Subcompartment Analysis (10) VI-8 Control Room Habitability 2.1.7.a & b Improved Auxiliary Feedwater System Reliability for PWRs a.
Automatic Initiation of the Auxiliary Feedwater System b.
Auxiliary Feedwater Flow Indication to Steam Generators NNECO Response These items do not apply to a BWR.
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e 2.1.8.a Instrumentation to Follew the Course of an Accident.
a.
Improved Post-Accident Sampling Capability NNECO Response The implementation criteria is being developed and as previously agreed upon will be submitted by the owners group on November 15, 1979.
In accot ance with the philorophy presented in the response to Item 2.1.6.b, procedural improvements or modifications which are judged to be independent of SEP topics will be implemented without consideration to the SEP schedule.
However, potential major modifications which may be required will be deferred until an integrated assessment of the related SEP topics, identified in 2.1.6.b, is completed.
2.1.8.b Instrumentation to Follow the Course of an Accident.
b.
Increased Range of Radiation Monitors NNECO Response NNECO concurs with the owners group response (Ref. 1)
NNEC0 will implement the requirements of NRC pc.ition 2.1.8b, Items 1, 2, and 3 as modified in the NRC Region 1 Meeting on September 24, 1979 for high-range noble gas effluent Fonitors, high-range conteinment radiation monitors, and high-range effluent radiciodine and particulate sampling and analysis.
NNECO intends to implement this requirement by January 1, 1981 to the extent possible consistent with commercial availability of equipment and the scheduling constraints discussed in the forwarding letter.
In addition, NNECO will develop and implement procedures for estimating noble gas and radiciodine release rates if the existing effluent instrumentation goes off-scale.
NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in our forwarding letter.
2.1.8.c Instrumentation to Follow the Course of an Accident.
c.
Improved In-Plant Iodine Instrumentation NNECO Response NNECO concurs with the owners group response (Ref.1) i205 144
NNECO will provide equipment, training, and procedures for determining airborne iodine concentration throughout the plant under accident conditions in accordance with NUREG-0578 requirements.
NNEC0 will implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
2.1.9
" Analysis of Design and Off-Normal Transients and Accidents".
NNECO Response NNECO concurs with the owners group response (Ref.1)
NUREG-0578 Implementation Letter Requirements Relative to Containment Level, Pressure, and Hydrogen Monitoring NNECO Response NNECO ccncurs with the owners group response (Ref. 1)
NNECO intends to comply with this requirement by January 1, 1981 to the extent possible in consideration of the scheduling constraints discussed in our forwarding letter.
NUREG-0578 Implementation Letter Requirement Relative to Remotely Operated High Point Vents.
NNEC0 Response NNECO concurs with the owners group response (Ref. 1) and adds the following specific information pertaining to Millstone Unit 1:
NNECO Design Relief valves operable from control room:
Location - On main steam lines inside containment Number - 6 Safety Grade - Yes Vessel head vent valves operable from control room - Yes Normally open vessel head vent line to main steam - Yes Steam turbine driven RCIC - NO Steam turbine driven HPCI - N0 1205 145
2.2.1.a Improved Reactor Operations Command Function a.
Shift Supervisor Responsibilities NNEC0 Response NNECO will implement the intent of this requirement.
However, in order to clarify the meaning of the term " accident situation" in item 2.b of the staff's position in Appendix A of NUREG-0578, the requirement is interpreted as follows: The Shift Supervisor or Supervising Control Operator shall until properly relieved remain in the control room at all times whenever a site or general emergency has been declared, to direct the activities of control room operators.
NNECO intends to implement this requirement by January 1, 1980.
2.2.1.b Improved Reactor Operations Connand Function b.
Shif t Technical Advisor Discussion Implementation of the Shift Technical Advisor (STA) as proposed by the Lessons Learned Task Force would place a graduate engineer independent and detached from plant operations in the control room at or shortly following the occurrence of an accident or abnormal transient.
Because the STA would not be in the direct operational chain of comand and, in fact, would not need to be licensed, he could neither manipulate nor direct licensed operators to manipulate the controls of the reactor plant.
He would be empowered to advise operations but not responsible to operations for his advice.
The Shift Supervisor is correctly charged with the responsibility for safe operation of the plant at all times. During the early phase of an accident, he discharges this responsibility by coordinating and directing the response of the control room staff. The actions of the operators are procedural, being governed by their training and emergency procedures, and during this phase the entire control room staff including the Shift Supervisor is completely occupied with responding to the accident.
Plant operating experience indicates that there is a period of time following initiation of any accident or transient wherein the shift supervisor has sufficient time to analyze, diagnose, and respond to the condition of the plant but does not have sufficient time to carefully consider an independent assessment of the accident, resolve any conflicts between his and the independent assessment and, on the basis of such assessment, decide to alter the procedural actions of the operators.
Dialogue regarding such an assessment or time spent resolving such conflicts can only distract and delay the Shift Supervisor and consequently degrade the respcnse of the control room staff to the accident.
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Even though the roles of shift superviser and STA can be carefully delineated by procedure and training, industrial and military experir se indicates that a direct-line organization wherein authority and re consibility are interdependent is required to effectively operate in a crisis environment.
The proposed STA is empowered to advise operations but not responsible to operations for his advice.
His authority and responsibility are not interdependent.
A potential for conflict and confusion exists which cannot be completely eliminated by procedure or training because procedure and training can address only those event sequences which have been postulated in advance. One important lesson learned from the experience at Three Mile Island and at other facilities is that not all event sequences can be postulated in advance.
Therefore, an alternative which avoids this potential for conflict and confusion but improved the functions intended by the proposed STA is recommended.
Two functions are intended to be improved by the proposed STA:
(1) accident assessment and (2) operating experience assessment.
In order to improve the accident assessment function while avoidiag the degradation in accident response which accompanies the proposed STA, the course of an accident is considered in three sequential phases:
immediate, intermediate, and recovery.
The immediate phase extends from the point at which an abnormal condition affecting plant safety can be detected in the control room until the point at which the Shift Supervisor has sufficient time to carefully consider an independent assessment and, on the basis of such assessment, decide to alter the procedural actions of the operators.
The intermediate phase extends from the end of the immediate phase until the point at which the Technical Support Center (TSC) is manned and ready.
The recovery phase extends from the end of the intermediate phase until the point at which recovery is complete.
For the immediate phase, the accident assessment function can be improved only by upgraded training to enhance the operators' abilities to recognize, diagnose, and respond to accident conditions.
During this phase the operators' actions are governed by training and emergency procedures, and by definition there is insufficient time for the careful consideration of an independent assessment which would be required before such an assessment could become the basis for altering the procedural actions of the operators.
For the intermediate phase, the accident assessment function can be improved by either of two alternative means.
An operator can be educated in science and engineering in order that he might provide an assessment which could be considered and acted upon by the Shift Supervisor.
Alternatively, a graduate engineer or equivalent can be trained in piant operations and made available to the Shift Supervisor en call in order that he might provide such an assessment.
In either case, the Shift Supervisor must have sufficient time to carefully consider the assessment and, based on such assessment, decide to alter the procedural actions of the operators.
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For the recovery phase, the accident assessment function can be improved by manning the TSC.
The collective engineering resource within the TSC will be able to develop a detailed independent assessment of plant con-ditions and provide appropriate procedures with which to recover from the accident.
The operating experience assessment function can best be provided by a team which reviews operating experience at the plant and at plants of like design.
Varying team membership as appropriate to the operating experience being assessed assures accomplishment of this function by the best qualified individuals.
NNECO Response The two functions intended to be improved by the proposed STA will be implemented as follows:
1.
Accident Assessment:
a.
Immediate Phase An operator or supervisor in the direct operational chain of command on each shift (normally in charge in the control room) will receive additional specific training in the response and analysis of the plant for transients and accidents.
This training will be coordinated with the schedule for preparation and review of analysis and guidelines under the NRC Bulletins and Orders Task Force.
All operators and supervisors wjl1 receive additional training appropriate to their fesponsibilities in the response of the plant to transients and accidents.
This longer term training qualification criteria will be provided by the Institute of Nuclear Power Operations.
b.
Intermediate Phase NNECO intends to implement the intermediate phase requirement for additional accident assessment capability by adding an additional licensed operator to each shift and upgrading the training of a senior reactor operator on each shift to include the general technical education, and additional transient and accident response training requirements as discussed in Enclosure (2) to D. G. Eisenhut's letter of September 13, 1979.
The senior reactor operator on each shift designated as the Shift Technical Advisor would have no assigned line functions while performing the Shift Technical Advisor function.
The addition of a licensed reactor to the control room shift compliment will crsble the STA to be detached from controls manipulation 1205 148
and supervision of operators during an evn t This provision is considered critical to our intended method of implementation since the primary deficiency noted in the staff's discussion regarding this altervative was the need for involvement of each of the current shift complement of three operators in satisfying the demands for prompt control and supervising actions.
Implementation of the STA requirement in this way also prevents dilution of command authority during an accident situation which was noted as not desirable in Enclosure (2) to D. G. Eisen-hut's letter of September 13, 1979.
NNECO feels that this method represents the optimum alternative for implementation of the STA requirement.
However, as discussed in Enclosure (2) referenced above, the completion of the additional general technical education and transient and accident training requirements may take two years or more to fully implement depending on the scope and content of the training requirements as finally established.
Therefore, in the interim period while the designated operators are off shift or offsite receiving the required training, NNECO intends to implement an interim method of providing the additional accident assessment capability.
The interim method which NNECO intends to implement by January 1, 1980 is to provide immediate on call assistance to the control room by designated senior perionnel from the plant staff available onsite in approximtately 30 minutes.
The majority of designated individuals will have a bachelors degree in science er engineering and all will have a current senior reactor operator license on the designated unit.
These individuals will also receive short term supplemental training and retraining in plant transient and accident response.
As an additional interim measure, NNEC0 will also provide an on call group of experts with experience and technical background in the various technical areas important to safety including mechanical, electrical and fluid systems, reactor physics, chemistry, and metallurgy.
This group would consist of individuals from both the plant staff and our Northeast Utilities Service Compa1y (NUSCO) engineering staff.
Due to the close geographical proximity of the NUSCO engineering support group, this on call team of experts will be available on cite within approximately one hour of an event.
c.
Recovery Phase Individuals knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident will be available on call to staff the On-Site S cnnical Support Center.
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2.
Operating Experience Assessment A team will be designated by NNEC0/NUSCO to assess the operating experience at Millstone Unit 2 and at plants of like design.
Team membership may vary as appropriate to the operating experience being assessed but will include degreed engineers with experience in the various technical areas important to safety including mechanical, electrical and fluid systems engineering, reactor physics, chemistry and metallurgy.
In addition the team membership will include or provide routine access to persons experienced in the principles of human engineering or human factors.
Procedures will be provided to insure close coupling of the results of the operating experience assessment function with the STA function.
The operating experience assessment function will be implemented by January 1, 1980.
- 2. 2.1. c Improved Reactor Operations Command Function c.
Shift and Relief Turnover Procedures NNECO Response NNECO will review and revise plant procedures as necessary to comply with this requirement.
NNECO intends to implement this requirement by January 1, 1980.
2.2.2.a Improved In-Plant Emergency Procedures and Preparations a.
Control Room Access NNECO Response NNECO will review and revise plant emergency procedures as necessary to comply with this requirement.
NNECO intends to implement this requirement by January 1, 1980.
2.2.2.b Improved In-Plant Emergency Procedures and Preparations b.
Onsite Technical Support Center NNECO Response NNECO will implement the requirement for an interim Technical Support Center in accordance with revised NUREG requirements as follows:
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1.
A location will be designated in the emergency plan.
This may be a temporary location.
2.
Telephone communications will be established to the control room and the NRC.
These may be temporary.
3.
The staffing and activation criteria will be specified in the emergency plan.
4.
The TSC will have access to plant technical information.
NNECO will submit a preliminary design package for the final TSC by January 1, 1980.
2.2.2.c Improved In-Plant Emergency Procedures and Preparations c.
Onsite Operational Support Center NNECO Response NNECO intends to comply with the requirements for this recommendation by January 1, 1980.
2.2.3 Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability NNECO Response Implementation of this item has been deferred by the NRC.
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DOCKET No. 50-336 ATTACH 3ENT 3 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 FOLLOWUP ACTIONS RESULTING FROM THE NRC STAFF REVIEWS REGARDING THE THREE MILE ISLAND, UNIT 2 ACCIDENT OCTOBER, 1979 1205 152
MILLSTONE NUCLEAR POWER STATION, 'JNIT NO. 2 FOLLOWUP ACTIONS RESULTING FROM T3E NRC STAFF REVIEWS REGARDING THE THREE MILE ISLANJ, UNIT 2 ACCIDENT Item 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief and Block Valves, and Pressurizer Level Indicators in PWR's.
NNECO Response The redundant emergency power requirements for pressurizer heaters, power-operated relief valves and associated block valves, and pressurizer level indication instrument channels will be met on Millstone Unit No. 2 in conformance with requirements appropriate for this unit. The functional requirements for the specified pressurizer subsyr,tems needed in order to provide sufficient availa-of a loss of offsite bility of the pressurizer for pressure control in the event These functional require-power are being developed by the NSSS Owners Group.
ments will be submitted for the Implementation Review no later than January 1, 1980. NNECO intends to implement these requirements by January 1, 1980.
Item 2.1.2 Perforcance Testing for BWR and PWR Relief and Safety Valves.
NNECO Response A program for testing power operated relief valves (PORV's) and safety valves (SV's) used for primary system pressure control under design bases operating conditions is being developed by the NSSS Owners Group.
This program includes definition of test conditions and qualification requirements for all specified The results of this program will be made available valves in operating reactors.
to the generic efforts being undertaken by the industry (thrc"gh for example the Electric Power Research Institute, EPRI, and the Nuclear Safety Analysis Center, NSAC) no later than January 1, 1980. These results will also be available for discussions with the NRC Staff to establish Generic Resolutions no later than January 1, 1980. Millstone Unit No. 2 will comply with the schedule for completion of the test program which is agreed to during these Generic Resolution meettngs.
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Item 2.1.3 Information to Aid Operators in Accident Diagnosis and Control, a) Direct !ctication of Power 0perated Relief Valve and Safety Valve Po.. tion f or PWR's and BWR's.
NNECO Response The Millstone Unit No. 2 primary system relief and safety valves will be provided with a position indication in the control room derived from either a reliable valve position indication device or a reliable indication of flow in the discharge pipe. The functional requirements and conceptual design will be submitted for the implementation Review no later than January 1, 1980.
NNECO intends to implement this requirement and submit the design details to the NRC by January 1,1980, to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.3 Information to Aid Operators in Accident Diagnosis and Control.
b)
Instrumentation for Detection of Inadequate Core Cooling for PWR's and BWR's.
NNECO Response The procedures to be used by an operator to recognize inadequate core cooling will be developed based on analyses being performed as required by Item 2.1.9, Transient and Accident Analysis, Analysis of Inadequate Core Cooling; these analyses are described below. The guidelines for the procedures are being developed by the NSSS Owners Group and will be avellable for discussions with the NRC Staff to ectablish Generic Resolutions no later than January 1, 1980.
If the analyses or the guidelines indicate the need for the design of new instrumentation, the design of such instrumentation will be made available for discussions with the NRC Staff to establish Generic Resolutions.
The functional requirements and a conceptual design for a reactor vessel level This effort measurement device are being developed by the NSSS Owners Group.
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. includes a survey of currently available technology and asseesment of the f easf bility of various alternatives.
If required by the Generic Resolution discussions with the NRC Staf f, the functional requirements and conceptual design will be submitted for Proposal Review by the NRC Staf f prior to implementation.
The installation schedule for such a device, should it be deemed necessary, will be established during the Proposal Review.
NNECO will install a safety grade primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that it is not to be used exclusive of other related plant parameters.
NNECO intends to implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.4 Containment Isolation Provisions for PWR's and BWR's.
NNECO Response The Millstone Unit No. 2 containment isolation system will comply with the recommendations of SRP 6.2.4 regarding diversity in the parameters sensed for the initiation of containment isolation. A generic review of all systems pene-trating the containment on all operating plants is being conducted by the NSSS Owners Group. This review will produce generic criteria for the definition of essential systems, identification of all such systems, and specification of the bases for each system's selection. Criter$a are also being developed for selective unisolation of non-essential systems which may be beneficial. The results of this review will be submitted for Implementatioa Review no later than January 1,1980.
NNECO intends to implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
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. Item 2.1.5 Post-Accident Hydrogen Control Systems for PWR and BWR Contain=ents.
a) Dedicated Penetrations for External Recombiner or Post-Accident External Purge System b) Inerting BWR Containments c) Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant NNECO Response These requirements are not applicable since Millstone Unit No. 2has hydrogen recombiners installed inside the containment.
It is also noted that the NRC has delayed proposed rule =aking proceedings on Ite=s 2.1.5b and 2.1.5c, requiring no action on these items at this time.
Item 2.1.6 Post-Accident Control of Radiation in Systems Outside Containment of PWR's and BWR's, a)
Integrity of Systems outside Containment Likely to Contain Radioactive Materials (Engineered Saf ety Systems and Auxiliary Systems)
NNECO Response NNECO will develop and implement a program to reduce leakage from syste=s outside containment that would or could contain high-level radioactive materials during a serious transient or accident to as low as practical levels. The program will include initial Onplementation of all practical leak reduction measures on the af fected systems and measurement of actual leakage rates with the systems in operation. A continuing program will also be established to implement preventative maintenance to reduce leakage and periodic leak testing at a frequency not to exceed refueling outage intervals.
The systems to which this program applies will be determined based upon the system function during and af ter a serious transient or accident. NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
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. is noted that this ef fort was initiated as part of *.he power uprating It process for Millstone Unit No. 2, and a license amendment request has been submitted as required by the Staff.
Item 2.1.6 Post-Accident Control of Radiation in Systems Outside Containment of PWR's and BWR 's.
b) Design Review of Plant Shielding of Spaces for Post-Accident Operations.
NNECO Response NSECO will perform a radiation and shielding design review of the spaces around systems which may contain highly radioactive materials as the result of an accident. The review will identify the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these The design review shall determine the corrective actions needed for systems.
vital areas throughout the plant.
Radiation level guidance and shielding source term criteria for this design review will be in accordance with the information presented by the NRC at the Region 1 Meeting on September 24, 1979.
NNECO intends to implement this requirement by January 1,1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter. The schedule for implementation of any required plant design modifications as a result of this design review will require further definition pending final determination of the required modifications.
It em 2.1. 7 Improved Auxiliary Feedwater System Reliability for PWR's a) Automatic Initiation of the Auxiliary Feedwater System NNECO Response Automatic initiation of auxiliary f eedwater is considered to be unnecessary f or Millstene Unit No. 2 for the following reasons:
(1) immediate actions are (2) required by the reactor trip procedure to verif y f eedwater flow status; there is complete control of the auxiliary feedwater system from the main 1205 157
. control board; (3) there is in excess of fif teen minutes available before auxiliary feedwater is required; and (4) past experience with recovery from f eedwater system problems indicates no need for automation of the auxiliary Each of these f acts was presented in detail to the Staf f feedwater system.
in the responses to the TMI-related I6E Bulletins and follow-up correspondence.
This resulted in the current operating procedures which require one of the two control room operators to immediately initiate auxiliary feedwater if required.
it is not credible to assume that auxiliary feedwater will not be initiated within 15 minutes if it is required. A program to refine these calculations and document the time available before dryout of the Millstone Unit No. 2 steam generators following total loss of feedwater flow from full power conditions is being conducted by the NSSS Owners Group. The results of this program will be submitted for Proposal Review by the NRC Staf f no later than January 1,1980, to further support this position, it is also noted tha t several other aspects of the issue of automatic initiation of auxiliary feedwater should be considered in the Staff evaluation. Millstone Unit No. 2 has a license condition restricting the flow of feedwater under certain conditions due to water hammer concerns.
NNECO submitted a request, with justification, to delete this license condition nearly one year ago, but it has yet to be dis-positioned by the Staff.
Secondly, automatic initiation of auxiliary f eedwater has the potential to complicate the feedwater line break event by increasing the coolaown rate of the primary system and increasing the severity of the environment within containment. The time required to address the above concerns renders the commitment to install automatic initiation particularly premature.
It em 2.1. 7 Improved Auxiliary Feedwater System Reliability for PWR's.
b) Auxiliary Feedwater Flow Indication to Steam Generators NNECO Response The Millstone Unit No. 2 auxiliary feedwater system will be provided with a control grade indication of flow to each steam generator in the control room.
The functional requirements for such an indication including the number of channels, range, control functions, alarm functions, display, recorded outputs, 1205 158
and setpoints are being developed by the NSSS Owners Group. These functional requirements will be submitted for Implementation Review by the NRC Staff no later than January 1,1980. NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints noted in our forwarding letter.
The Millstone Unit No. 2 auxiliary feedwater system will be provided with a safety grade indication of flow to each steam generator in the control room.
The functional requirements for such an indication are being developed by the NSSS Owners Group. These functional requirements will be submitted for Implementation Review by the NRC Staff no later than January 1, 1981.
NNECO intends to implement this requirement by January 1, 1981 to the extent possible in consideration of the scheduling constraints noted in our forwarding letter.
Item 2.1.8 Instrumentation to Follow the Course of an Accident a)
Improved Post-Accident Sa=pling Capability NNECO Response NNECO will perform design and operational reviews of the reactor coolant and containment atmosphere sampling systems, radiological spectrum analysis facilities, and chemical analyses capabilities in accordance with the require-ments of NUREG-0578. NNECO intends to implement these requirements by January 1, 1980, to the extent possible in consideration of the scheduling constraints noted in the forwarding letter.
If the review indicates that the required sampling or analyses cannot be performed in a prompt manner using existing equipment, design modifications will be developed and proposed to the NRC by January 1, 1980. The implementation schedule for any required design modifications will be addressed in the January 1, 1980 submittal.
Item 2.1.8 Instrumentation to Follow the Course of an Accident b)
Increased Range of Radiation Monitors
NNECO Response NNECO will implement the requirements of NRC position 2.1.8b, Items 1, 2, and 3 as modified in the NRC Region 1 Meeting on September 24, 1979 for high-range noble gas effluent monitors, high-range containment radiation monitors, and high-range effluent radioiodine and particulate sampling and analysis.
NNECO intends to implement this requirement by January 1,1981 to the extent possible consistent with commercial availability of equipment and the scheduling constraints discussed in the forwarding letter.
In addition, NNECO will develop and implement procedures for estimating noble gas and radioiodine release rates if the existing effluent instrumentation goes off-scale.
NNECO intends to implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in our forwarding letter.
Item 2.1.8 Instrumentation to Follow the Course of an Accident c)
Improved In-Plant Iodine Instrumentation NNECO Response NNECO will provide equipment, training, and procedures for determining air-borne iodine concentration throughout the plant under accident conditions in accordance with NUREG-0578 requirements. NNECO will implement this requirement by January 1, 1980 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.1.9 Analysis of Design and Off-Normal Transients and Accidents NNECO Response The response to Transient and Accident Analysis requirements is being developed by the NSSS Owners Group in conjunction with General Resolution meetings with the NRC Bulletins and Orders Task Force. These responses will be sub=itted on the schedule agreed to by that Task Force and the NSSS Owners Group and will 1205 160
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be referenced for specific application. The implamentation of emergency procedures and retraining will be done by NNECO on a schedule consistent with that established for the analysis requirements.
Additional Instrumentation, Containment Pressure, Containment Water Level and Hydrogen Monitors, to follow the Course of the Accident NNEC0 Response NNECO intends to comply with this requiremera by January 1, 1981 to the extent possible in consideration of the scheduling constraints discussed in our forwarding letter.
Installation of Remotely Operated High Point Vant in Reactor Coolant System NNECO Response NNECO will implement this requirement in accordance with the criteria discussed at the October 11, 1979 NRC Topical Meeting on this subject.
NNECO intends to implement this requirement by January 1, 1981 to the extent possible in consideration of the scheduling constraints discussed in the forwarding letter.
Item 2.2.1 Improved Reactor Operations Command Function a)
Shift Supervisor Responsibilities NNECO Response NNECO will implement the intent of this requirement. However, in order to clarify the meaning of the term " accident situation" in Item 2.b of the Staff's position in Appendix A of NUREG-0578, the requirement is interpreted as follows:
The Shift Supervisor or Supervising Control Operator shall until properly relieved remain in the control room at all times whenever a site or general emergency has been declared, to direct the activities of control room operators. NNECO intends to implement this requirement by January 1, 1980.
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Iten 2.2.1 Improved Reactor Operations Command Function b) Shif t Technical Advisor Discussion Implementation of the Shif t Technical Advisor (STA) as proposed by the Lessons Learned Task Force would place a graduate engineer independent and detached f rom plant operations in the control room at or shortly following the occurrence of an accident or abnormal transient.
Because the STA would not be in the direct operational chain of command and, in f act, would not need to be licensed, he could neither manipulate nor direct licensed operators to manipulate the controls of the reactor plant.
He would be empowered to advise operations but not responsible to operations for his advice.
The Shift Supervisor is correctly charged with the responsibility for saf e operation of the plant at all times. During the early phase of an accident, he discharges this responsibility by coordinating and directing the response of the control room staff. The actions of the operators are procedural, being governed by their training and emergency procedures, and during this phase, the entire control room staf f including the shif t supervisor is completely occupied with responding to the accident. Plant operating experience indicates that there is a period of time following initiation of any sccident or transient wherein the Shift Supervisor has suf ficient time to analyze, diagnose, and respond to the condition of the plant but does not have suf ficient time to carefully consider an independent assessment of the accident, resolve any conflicts between his and the independent assessment and, on the basis of such assessment, decide to alter the procedural actions of the operators. Dialogue regarding such an assessment or time spent resolving such conflicts can only distract and delay the Shif t Supervisor and consequently degrade the response of the control room staff to the accident.
Even though the roles of Shif t Supervisor and STA can be carefully delineated by procedure and training, industrial and military experience indicates that a direct-line organization wherein authority and responsibility are inter-dependent is required to ef fectively operate in a crisis environment. The proposed STA is empowered to advise operations but not responsible to operations for his advice. His authority and responsibility are not inter-dependent. A potential for conflict and confusion exists which cannot be completely eliminated by procedure or training because procedure and training can address only those event sequences which have been postulated in advance.
One important lesson learned from the experience at Three Mile Island and at other facilities is that not all event sequences can be postulated in advance.
Therefore, an alternative which avoids this potential for conflict and confusion but improves the functions intended by the proposed STA is recommended.
Two functions are intended to be improved by the proposed STA:
(1) accident assessment and (2) operating experience assessment.
In order to improve the accident assessment function while avoiding the degradation in accident response which accompanies the proposed STA, the course of an accident is considered in three sequential phases:
immediate, intermediate, and recovery.
1205 162 The immediate phase extends from the point at which an abnormal condition affecting plant saf ety can be detected in the control room until the point at which the Shif t Supervisor has suf ficient time to carefully consider an independent assessment and, on the basis of such assessment, decide to alter the procedural actions of the operators. The intermediate phase extends from the end of the immediate phase until the point at which the Technical Support Center (TSC) is manned and ready. The recovery phase extends from the end of the intermediate phase until the point at which recovery is complete.
For the immediate phase, the accident assessment function can be improved only by upgraded training to enhance the operators' abilities to recognize, diagnose, and respond to accident conditions. During this phase, the operators' actions are gccerned by training and emergency procedures, and by definition, there is insuf ficient time for the careful consideration of an independent assessment which would be required before such an assessment could become the basis for altering the procedural actions of the operators.
For the intermediate phase, the accident assessment function can be improved by either of two alternative means. An operator can be educated in science and engineering in order that he might provide an assessment which could be considered and acted upon by the Shif t Supervisor.
Alternatively, a graduate engineer or equivalent can be trained in plent operations and made available to the Shif t Supervisor on call in order that he might provide such an assessment.
In either case, the Shift Supervisor must have sufficient time to carefully consider the assessment and, based on such assessment, decide to alter the procedural actions of the operators.
For the recovery phase, the accident assessment function can be improved by manning the TSC.
The collective engineering resource within the TSC will be able to develop a detailed independent assessment of plant conditions and provide appropriate procedures with which to recover from the accident.
The operating experience assessment function can best be provided by a team which reviews operating experience at the plant and at plants of like design.
Varying team membership as appropriate to the operating experience being assessed assures accomplishment of this function by the best qualified indi-viduals.
NNECO Response The two functions intended to be improved by the proposed STA will be implemented as follows:
(1) Accident Assessment a.
Immediate Phase An operator or supervisor in the direct operational chain of command on each shif t (normally in charge in the control room) will receive additional specific training in the response and analysis of the plant for transients and accidents. This training will be coordinated with 1205 163
. the schedule for preparation and review of analysis and guidelines under the NRC Bulletins and Orders Task Force.
All operators and supervisors will receive additional training appropriate to their responsibilities in the response of the plant to transients and accidents. This longer term craining and qualification criteria vill be provided by the Institute of Nuclear Power Operations.
b.
Intermediate Phase NNECO intends to implement the intermediate phase requirement for additional accident assessment capability by adding an addi-tional licensed operator to each shif t and upgrading the training of a senior reactor operator on each shif t to include the general technical education, and additional transient and accident response training requirements as discussed in Enclosure (2) to D. G. Eisenhut'r letter of September 13, 1979. The senior reactor operator on each shif t designated as the Shif t Technical Advisor would have no asaigned line functions while performing the Shif t Technical Advisor function.
The addition of a licensed reactor operator to the control room shift complement will enable the STA to be detached from controls manipulation and supervision of operators during an event. This provision is considered critical to our intended method of implementation since the primary deficiency noted in the Staf f's discussion regarding this alternative was the need for involvement of each of the current shif t complement of three operators in satisfying the demands for prompt control and supervisory actions.
Implementation of the STA requirement in this way also prevents dilution of command authority during an accident situation which was noted as not desirable in Enclosure (2) to D. G. Eisenhut's letter of September 13, 1979.
NNECO feels that this metted represents the optimum alternative for implementation of the STA requirement. However, as discussed in Enclosure (2) referenced above, the completion of the additional general technical education and transient and accident training i205 164
requirements may take two years or more to fully implement depending on the scope and content of the training requirements as finally established. Therefore, in the interim period while the designated operators are of f-shif t or off-site receiving the required training, NNECO intends to Lsplement an interim method of providing the additional accident assessment capability.
The interim method which NNECO intends to implement by January 1, 1980 is to provide immediate on-call assistance to the control room by designated senior personnel from the plant staff available on-site in approximately 30 minutes. The majority of designated individuals will have a Bachelor's Degree in science or engineering and all will have a current Senior Reactor Operator License on the designated unit. These individuals will also receive short-term supplemental training and retraining in plant transient and accident response. As an additional interim measure, NNECO will also provide an on-call group of experts with experience and technical backgrounds in the various technical areas important to safety including mechanical, electrical, and fluid systems; reactor physics, chemistry, and metallurgy. This group would consist of individuals from both the plant staff and our Northeast Utilities Service Company (NUSCO) engineering staff. Due to the close geographical proximity of the NUSCO engineering support group, this on-call team of experts will be available on-site within approximately one hour of an event.
c.
Recovery Phase Individuals knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident will be available on-call to staff the On-Site Technical Support Center.
(2) Operating Experience Assessment A team will be designated by NNECO/NUSCO to assess the operating experience at Millstone Unit No. 2 and at plants of like design. Team membership may vary as appropriate to the operating experience being assessed but will 1205 165
. include degreed engineers with experience in the various technical areas important to safety including mechanical, electrical and fluid systems engineering, reactor physics, chemistry, and metallurgy-In addition, the team membership will include or provide routine access to persons experienced in the principles of hu=sn engineering or human factors.
Procedures will be provided to insure close coupling of the results of the operating experience assessment function with the SIA f unc tion. The operating experience aesessment function will be imple-mented by January 1,1980.
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. Item 2.2.1 Improved Reactor Operations Co==and Function c) Shift and Relief Turnover Procedures NNECO Response NNECO will review and revise plant procedures as necessary to comply with this requirement.
NNECO intends to implement this requirement by January 1, 1980.
Item 2.2.2 Improved In-Plant Emergency Procedures and Preparations a)
Control Room Access NNECO Response NNECO will review and revise plant em;rgency procedures as necersary to comply with this requirement.
NNEC^ Aatends to L:plement this requirement by January 1,1980.
Item 2.2.2 Improved in-Plant Emergency Pracedures and Preparations b) Onsite Technical Supp:rt Center NNECO Response NNECO will implement the requirement for an interim Technical Support Center, in accordance with revised NUREG requirements, as follows:
(1) A location will be designated in the emergency plan. This may be a temporary location.
(2) Telephone communications will be established to the control room and the NRC. These may be temporary.
(3) The staffing and activation criteria will be specified in the emergency plan.
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(4) The TSC will have access to plant technical informatisn.
NNECO intends to implement this requirement by January 1,1980 to the extent possible in con'i.deration of the scheduling constraints discussed in the for-warding letter.
NNECO will submit a preliminary design package for the final TSC by January 1, 1980.
Item 2.2.2 Improved In-Plant Dsergency Procedures and Preparations c) Onsite Operational Support Center NNECO Response NNECO in' ends to comply with the requirements for this recom=endation by January 1, 1980.
Item 2.2.3 Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Saf ety System Availability NNECO Response implementation of this requirement has been def erred by the NRC.
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