ML19208D675

Supplemental Rept, Response to NUREG-0578,App A,TMI-2 Lessons Learned Task Force Short Term Recommendations

ML19208D675
Person / Time
Site:	Black Fox
Issue date:	07/27/1979
From:	PUBLIC SERVICE CO. OF OKLAHOMA
To:
Shared Package
ML19208D673	List: ML19208D672 ML19208D675 ML19208D676
References
NUDOCS 7909290287
Download: ML19208D675 (54)

Text

, . >

RESPONSE OF PUBLIC SERVICE COMPANY OF OKLAHOMA BLACK FOX STATION, UNITS 1 & 2 USNRC DOCKET NOS. STN 50-556, 50-557 TO NUREG-0578, Appendix A TMI-2 Lessons Learned Task Force Short-Term Recommendations Inspection & Enforcement Bulletin 79-08 Selected Is!cas on Emergency Preparedness 1055 297 July 27, 1979.

7 9 0929 d,[{ chL 7

TABLE OF CONTENTS Page Introduction & Description of Methodology. . . . . . . . . . . . . . . . . . . 1 Response to NUREG-0578, Appendix A. . . . . . . . . . . . . . . . . . . . . . .

Section 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves t.nd Block Valves and Pressurizer Level Indicators in PWR's . . . . . . . . . . . . . . . . . . 4 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves. . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.1.3.a Direct Indication of Power-0perated Relief Valves and Safety Valve Position for PWR's and BWR's. . . . . . . . . . . . . 6 2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWR's and BWR's. . . . . . . . . . . . . . . . . . . . . . . 7 2.1.4 Diverse and More Selective Containment Isolation Provisions for PWR's and 9WR's. . . . . . . . . . . . . . . . . . . . . . . 9 2.1. 5. a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems. . . . . . . . . . . . . . . . . . . . . . . . 10 2.1.5.b Inerting BWR Containments. . . . . . . . . . . . . . . . . . . . 11 2.1.5.c Capability to Install Hydrogen Recombiner at Each Light Water N ucl e a r Powe r Pl ant . . . . . . . . . . . . . . . . . . . . . 12 2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWR's and BWR's. . . . . . . . . . . . . . . . 13 2.1.6.b Design Review of Plant Shielding of Spaces for Post-Accident Operations. . . . . . . . . . . . . . . . . . . . . . . . . 14 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWR's. . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generators for PWR's. . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 2.1.8.a Improved Post-Accident Sampiing Capability. . . . . . . . . . . 17-2.1.8.b Increased Range of Radiation Monitors. . . . . . . . . . . . . . 18 1055 298

Page Table of Contents Improved In-Plant Iodine Instrumentation. . . . . . . . . . . 19 2.1.8.c 2.1.9 Analysis of Design and Off-Normal Transients and 20 Acc i den ts . . . . . . . . . . . . . . . . . . . . . . . .

2.2.1.a Shift Supervisor's Responsibilities. . . . . . . . . . . . . 23 2.2.1.b Shift Technical Advisor. . . . . . . . . . . . . . . . . . . 25 2.2.1.c Shift and Relief Turnover. Procedures. . . . . . . . . . . . . 26 2.2.2.a Control Room Access . . . . . . . . . . . . . . . . . . . . . 27 2.2.2.b Onsite Technical Support Center. . . . . . . . . . . . . . . 28 2.2.2.c Onsite Operational Support Center. . . . . . . . . . . . . . 29 -

2.2.3 Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability. . . . . . . 30 Response to Inspection and Enforcement Bulletin 79-08. . . . . . . . . . . 32 Response to Selected Issues on Emergency Preparedness. . . . . . . . . . . 45 1035 299

INTRODUCTION AND DESCRIPTION OF METHODOLOGY On June 15, 1979, Public Service Company of Oklahoma (PS0) submitted an analysis of the lessons to be learned from the events at Three Mile Island-Unit 2 as they apply to the construction permit application for the Black Fox Station (BFS). The submittal was documentation of the Company's long-term corporate comitment to incorporate those lessons into the design, staffing, training and operation of BFS. In addition, the document represented the initial effort by the PS0 Technical Advisory Comittee (TAC) constituted by the President and Chief Executive officer as an ongoing body expressly to study the events at TMI and to implement the lessons learned into our project.

With the issuance on July 19 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recomendations," the TAC compared the 23 lessons learned with our submittal. Although our June 15 analysis addressed most of the issues discussed in NUREG-0578, we found the organization of the material to differ in form. Hence, we chose to reiterate our comitments herein in accordance with the format of Appendix A to NUREG-0578.

Prior to development of this document, consultants to and members of the Technical Advisory Comittee met on June 19 with appropriate members of the regulatory staff, including Mr. Varga, Mr. Thomas, Mr. Silver, Mr. Williams, to review the intent of the NUREG-0578 technical positions.

In study of the twenty-three issues, we found that three (2.1.1, 2.1.7a, 2.1.7b) did not apply to BFS because the issue was specific to pressurized water reactors.

Three others (2.1.5 a, b, c) were not applicable because of the design features of the Black Fox Station which utilizes the BWR/6 Mark III System. Finally, one issue (2.2.3) did not apply since it is to be the subject of rulemaking.

1055 300

For the balance, the intent of each comitment by PS0 is to meet the express position of the regulatory staff as stated in NUREG-0578, Appendix A.

During our meetings with the regulatory staff and the Director of Nuclear Reactor Regulation, Mr. Denton, on July 19 and 20, it became apparent that the BFS was expected to address itself to the activity of the Bulletins and Orders Task Force. In the meeting of June 20, Messrs. Novak and Kane of the B&O TF stated that the only issues that need to be addressed by the BFS were those contained in Inspection and Enforcement Bulletin (IEB) 79-08.

The June 15 submittal by PS0 was intended to incorporate all of the requirements statedin IEB 79-08. In order to be completely responsive, each of the IEB 79-08 Tasks are repeated in this submittal followed by the appropriate PS0 comitment for BFS.

The IEB 79-08 was specifically addressed to licensees with operating boiling water reactors and response was required very quickly. For projects such as BFS having yet to receive a full construction permit and where operation is projected well into the future, the requirements of IEB 79-08 were provided for information purposes. No written response was required, but actions will be completed prior to start of operation. The PS0 comitments to action require completion of the efforts described during final design as detailed in the FSAR and in subsequently developed operating procedures.

- PS0 recognizes that the " Lessons Learned" requirements and the IEB 79-08 requirements represent separate activities within the regulatory staff. Thus, there exists some duplication of subject matter with the possibility of different interpretations of the PS0 response between the two task forces. If such differences are identified, PS0 comits to work with the NRC Staff to reconcile then.

_2 1055 301

There are several issues related to the events at TMI which relate to radiological emergency planning. These are being evaluated by a NRC group headed by Mr. Brian Grimes who met with PS0 on July 20, 1979. Mr. Grimes identified six matters which PSO should address in this submittal. Most were covered in our June 15 assessment Included in the emergency preparedness section is a letter from the Governor of the State of Oklahoma, George Nigh to Joseph Hendrie, Chairman USNRC. Therein, the status of the State Emergency Response Plan, PS0's role in development, and a ccmmitment to have a NRC approved plan in effect well before BFS commerctal operation is discussed.

PS0 has also confirmed the feasibility of implementing a protective action plan over the area covered by a ten-mile radius from the BFS generation complex, a possible future licensing criteria mentioned by Mr. Grimes.

The PS0 Technical Advisory Committee concurs with the view presented during the meetings of July 19 and 20, that all of the comitments and acticns required by the NRC Staff can be satisfied during the post-construction permit phase of the Black Fox design and construction effort, and that the documentation of these activities should be set forth in the Final Safety Analysis Report and Station Operating Procedures for the Black Fox Station. Our commitments reflect this understanding and philosophy.

1055 302

RESPONSE TO NUREG-0578, Appendix A TMI-2 Lessons Learned Task Force Short-Term Recommendations 1055 303

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Emergency Power Supply Reouirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in Pk'rl's Section 2.1.1 .

This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.

1055 304

9 NRR Lessons Learned Task Force Short-Term Reconmendations TITLE: Performance Testing for BWR and PWR Relief and Safety Valves (Section 2.1.2).

NRC STAFF POSITION Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor cooling system relief and safety valves under expected operating conditions for design basis transients and accidents. The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The -

signal failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maxirhized. Test pressures shall be the highest predicted by conventional safety analyses procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and support as well as the valves themselves.

PS0 COMMITMENT PS0 believes that it is important to assure that the safety and relief valves installed in the BFS reactor coolant boundary will function as intended and maintain their integrity under expected operating conditions for design basis transients and accidents. Analysis of accidents and transients will be conducted during the final design stage to determine the most severe operating conditions and dynamic forces experienced by the safety and relief valves during the selected events. PSO, in cooperation with other applicants and licensees, will conduct necessary testing to qualify the reacter coolant system relief and safety valves for the most severe conditions identified.

Qualification of the associated control circuitry and piping and supports will be verified at the test conditions selected for the safety and relief valves.

Documentation will be contained in the FSAR at the time of submittal in support of the operating license application.

1055 305

N_RR tessons Learned Task Force Short-Term Recommendations TIT'.E: Direct Indication of Power-Operated Relief V&lve ard Safety Vaive--

Position for PWR's and BWR's Section 2.1.3.a .

NRC STAFF POSITION Reactor system relief and safety vElves shall be provided with a positir< Indi-cation in the control room derived from a reliable valve position detect: >n device or a reliable indication of flow in the discharge pipe.

PS0 COMMITMENT PS0 will provide a reliable safety and relief valve position indication in the control room for the nineteen reactor main steam safety / relief valves in each nuclear steam supply system. Design detail will be provided in the FSAR.

1055 306

NRR Lessons Learned Task Force Short-Term Reconrnendations TITLE: Instrumentation for Detection of Inadequate Core Coolino in PWR's and BWR's Section 2.1.3.b .

NRC STAFF POSITION

1. Licensees shall develop pro:edures to be used by the operator to recognize inadequate core cooling with currently available instrumentation.

The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions. A detailed description of the analyses needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement, " Analysis of Off-Normal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).

In addition, each PWR shall install a primary coolant saturation neter to provide on-line indication of coolant saturation condition. Operator instructions as to use of this meter shall include consideration that is not to be used exclusive of other related plant parameters.

2. Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description cf the functional design requirements for the system shall also be included.

A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

PS0 COMMITMENT The ability of station operators to easily and unambigucusly determine the status of core cooling and to provide adequate cooling is essential to the operation of the Black Fox Station. PS0 will review the instrumentation presently provided within the BFS design to assure that adequate information is available for the clear definition of core cooling status. Should modifications or additional instrumentation be required to provide operators with clear, easily interpreted information, appro-priate modifications or additions to instrumentation will be provided during final design. Operating procedures will be developed to guide the operator in recognizing inadequate core cooling, and operators will be throroughly trained in the procedure and utilization of instrumer.tation to assure correct interpretation.of the core

~7-1055 307

cooling status. A description of system functional requirements and of the instru-mentation provided to enable operators to evaluate core cooling will be presented in the FSAR.

O b

e 1055 308

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Containment Isolation Provisions for PWR's and BWR's (Section 2.1.4).

NRC STAFF POSITION

1. All containment isolation system designs shall comply with the recomendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.

2. All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs acccrdingly, and shall report the results of the tv-evaluation to the NRC.

3. All non-essential systems shall be automatically isolated by the containment isolation signal.

4. The design of control systems for automat a containment isolation valves shall be such that resetting the istlation signal will not result in the automatic reopening of containment isolation vi.1ves. Reopening of con-tainmentisolation valves shall rcquire deliberate Sarstnr action.

PS0 COMMITMENT _

PS0 recognizes the importance for t#:c'y and effective isolation of the containment under accident conditions. P50 will raview the design of BFS to assure that the final design provides for:

1. Diversity in the parameten sensed for the initiation of containment isolation, in accordance with SRP 6.2.4;

2. Automatic isolation of non-essential systems upon containment isolation signal;

3. Reopening of containment isolation valves only by deliberate operator action. The control system design will not cause the automatic reopening of containment isolation valves upon resettling of the isolation signal.

The definition of essential and non-essential systems will be re-evaluated to carefully identify essential systems and non-essential systems to assure that the bases for selection of essential systems are described, and that the containment isolation design is consistent with the definition. The results of the re-evaluation will be reflected in the final containment design as presented in the FSAR, including inforstion on the definition of essential and non-essential systems.

1055 309

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems Section 2.1.5.a .

NRC STAFF POSITION Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmostphere should provide containment isolation systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.

Black Fox Station is designed for the installation of 100% redundant hydrogen recombiners within the containmer.t of each unit. This position is therefore not applicable.

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Inerting GWR Containments (Section 2.1.5.b).

NRC STAFF POSITION It shall be required that the Vermont Yankee and Hatch 2 Mark I BWR contain-ments be inerted in a manner similar to other operating BWR plants. Inerting shall also be required for near term OL licensing of Mark I and Mark II BWR's.

Black Fox Station is designed.with a Mark III Containment. This position is not applicable.

1 U 55  !,i 1

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Capability to Install Hydrogen Recombiner at Each Light Water huclear Power Plant Section 2.1.5.c .

NRC STAFF POSITION (Minority View).

1. All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following r,n accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.

2. The procedures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering shielding requirements and personnel exposure limitations as demonstrated to be necessary in the case of Ti11-2. .

Black Fox Station is designed for the installation of 100% redundant hydrogen recombiners within the containment of each unit. This position is therefore not applicable to BFS.

1055 512

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Integrity of Systems Outside Containment Likely to Contain Radioactive -~

Materials Engineered Safety Systems and Auxiliary Systems for PWR's and BWR's Section 2.1.6.a).

NRC STAFF POSITION Applicants and licensees shall immediately implement a program to reduce leakage frem systems outside containment that wculd or could centain highly radioactive fluids during a serious transient or accident to as-low-as practical levels. This program shall inclu de the following:

1. Immediate Leak Reduction.

a. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.

b. Measure actual leakage rates with system in operation and report them to the NRC.

2. Continuing Leak Reduction.

Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.

9 PS0 COMMITMENT PS0 will perform a review during the course of final design and make changes accordingly to provide a means of practical leak detection in systems outside containment which could be expected to have highly radioactive fluids as a result of a serious transient or accident. The review will also examine methods of leak repairs to achieve ALARA. Prior to initial operations, a preventive maintenance program shall be implemented to control the leakage, including periodic integrated leak rate tests, at a frequency not to exceed the refueling cycle interval.

The FSAR will contain the results of the above design and operations review.

1055 3i3

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Design Review of Plant Shielding of Spaces for Post-Accident Operations Section 2.1.6.b .

NRC STAFF POSITION With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas -

and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers,md instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by tne radiation fields during post-accident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

PS0 COMMITMENT PSO recognizes, as a result of the TMI-2 event, the need to assure necessary access to vital areas and protection of vital equipment under the impact of post-accident releases of radioactivity. PS0 will identify vital areas and equipment, and based on the post-accident radioactivity releases describcd in Regulatory Guide 1.3, will evaluate the BFS design for unacceptable limitations on personnel access and occupancy or undue degradation of safety-related equipment during post-accident operations. The evaluation will consider alternatives, including layout changes, increased use of permanent shielding, temporary shielding, or procedural controls.

The evaluation will determine changes needed throughout Black Fox Station. The results of the evaluation and a description of the changes will be reflected in the final design presented in the FSAR.

1055 314

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Automatic Initiation of the Auxiliary Feedwater System for PWR's 6ection 2.1.7.a).

This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.

1055 315

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Auxiliary Feedw ter Flow Indication to Steam Generators for PWR's Section 2.1.7.b .

This issue is not applicable to the BWR/6 Nuclear Steam Supr ' y System of the Black Fox Station, Units 1 and 2.

1055 516

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Imoroved Post-Accident Samoling Capability (Section 2.1.8.a).

NRC STAFF POSITION A design and operational review of ~the reactor coolant and containment atmos-phere sampling systems shall be performed to detennine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 rems to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.

If the review indicates that personnel could not promptly and safety obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly (less.than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) quantify certain radioisotopes that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures). and non-volatile isotopes (which indicate fuel metling). The initial rerctor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.

The review should also consider the effects of direct radiation froa piping and components in the auxiliary building and possible cor aminai. ion and direct radiatien from airborne cffluents. If the review indicates that the analyses required cannot be perforemd in a prompt manner with existing eqaipment, then desigr. modifications or equipment procurement shall be undertaken to meet tne criteria.

Ir; addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall ba capable of being completed promptly; i.e, the boron sample analysis within an hour and the chloride sample analysis within a shift.

PS0 COMMITMENT PS0 will perform a design and operational review of the reactor coclant and con-tainment atmospheric sampling system, the radioisotope analysis facilities, and chemical analyses to achieve prompt and safe sampic acquisition and analysis in accordance with the position stated above. Results of these studies will be presented in the FSAR.

1 CSS 517

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Increased Range of Radiation Monitors (Section 2.1.8.b).

NRC STAFF POSITION The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident,"

which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term.

1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal ceprating conditior.s; multiple monitors are considered to be necessary to cover the ranges of interest.

a. Noble gas effluent monitors with an upper range capacity of 105 uCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

b. Noble gas effluent monitoring shall be provided for the total range of concentration extending from a minimum of 10-7 uCi/cc (Xe-133) to a maximum of 105 uCi/cc (Xe-133). Multiple monitors are considered to be accessary to cover the ranges of interest. The range capacity of individual monitors shall overlap by a factor of ten.

2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiciodines for the accident condition shall be pruvided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.

J. In-containment radiation 1cvel monitors with a maximum range of 10 8 rad /hr shall be installed. A minimum of two such monitors that are physically separatedshall be provided. Monitors shall be designed and qualified to function in an accident environment.

pSO COMMITMENT PS0 shall provide the monitors as required in the staff position, and will document a description of the same in the FSAR.

1055 ',18 NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Improved In-Plant Iodine Instrumentation (Section 2.1.8.c).

NRC STAFF POSITION Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions.

PS0 COMMITMENT .

PS0 will provide instrumentation, training of personnel ard the technical procedures for accurately determining airborne iodine concentration throughout the plant under accident conditions, with documentation to be provided in the FSAR.

1055 319

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Analysis of Design and Off-Normal Transients and Accidents Section 2.1.9 .

NRC STAFF POSITION Analyses, procedures, and training addressing the following are required:

1. Samil break loss-of-raolant accidents;

2. Inadequate core cooling; and .

3. Transients and accidents.

Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force. These should be completed. In addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests, (scheduled to start in September,1979) shall be performed as a means to verify the analyses performed in support of the small break emergancy proce-dures and in support of an eventual long-term verification of compliance with Appendix K of 10 CFR Part 50.

In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:

1. Low reactor coolant system inventory (two examples will be required:

LOCA with fcrced flow; LOCA without forced flow);

2. Loss of natural circulation (due to loss of heat sink).

These calculations shall include the period of time during which inadequate core cooling is apprcached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included.

Each case should then be repeated taking credit for correct operator action.

These additional cases will provide the basis for developing appropriate emergency procedures. These calculations should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3b in this appendix).

The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR. The analyses shall include a single active failure for each system called upon to function for a particular event.

Consequential failures shall also be considered. Failures of the operators te perform required control manipulations shall be given consideration for permutations of the analyses. Operator actions that could cause the complete loss of function of a safety system shall also be considered. At preser t, these analyses need not address passive failures or multiple system failures in the short term. In the recent analysis of small break LOCA's, complete loss of auxiliary feedwater was considered. The complete loss of auxiliary feedwater j p q r, 320

Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--

Continued.

may be added to the failures being considered in the analysis of transients and accidents if it is concluded that more is needed in operator training beyond the short-term actions to upgrade auxiliary feedwater system reliability.

Similarly, in the long tem, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-term analyses.

The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations could be used to provide important quantitative information not available from an event tree.

For example, failure to initiate high-pressure injection could lead to core uncovery for some transients, and a computer calculation could provide information on the amount of time available for corrective action. Reactor simulators may provide some information in defining the event trees and would be useful in studying the information available to the operators. The transient and .

accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of Ere serious accidents.

The information derhed from the preceding analyses shall be included in the plant emergency procedures and operator training. It is expected that analyses performed by the NSSS vendors will be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant.

In addition to analyses performed by the reactor vendors, analyses of selected transients should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor vendors. These comparisons together with comparisons to data, including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures.

PS0 COMMITMENT As the penultimate paragraph of the above stated position of the NRC staff indicates, the requirement for additional transient and accident analyses is promoted by the need to develop more knowledge and information for reactor operations rather than a concern about the adequacy of reactor design. Information of this type is best developed on a generic basis, and as indicated below, such information will be available prior to the operation of the Black Fox Station.

PS0 understands that analysis and emergency procedures or guidelines for:

1. Small break loss-of-coolant accidents;

2. Inadequate core cooling; and 1055 321 Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--

Continued.

3. Transients and accidents are being generated by the operating Boiling Water Reactor Owners' Group in response to the Bulletins and Order Task Force. These analyses are being generalized first to cover BWR/1-5 type power plants and will be extended by General Electric Company to cover the BWR/6 System generically. Each of the specific requirements stated in the 6bove position have been identified by the Bulletins and Orders Task Force. As this assessment is completed for the operating power plants, the results will be reflected in the FSAR and factored into the Black Fox Station plant emergency procedures development and operator training. Analyses performed by General Electric will be put in the form of emergency procedures guidelines, and these guidelines will be implemented in the Black Fox Station procedures and training programs as appropriate.

1055 322

!1RR Lessons Learned Task Force Short-Term Recommendations TITLE: Shift Supervisor's Responsibilities (Section 2.2.1.a).

NRC STAFF POSITION The highest level of corporate management of each licensee shall

~

1.

issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions on his shift and that clearly establishes his comand duties.

2. Plant procedures shall be reviewed to assure that the duties, responsi-bilities, and authority of the shift supervisor and control rcom operators are properly defined to effect the establishment of a definite line of command and clear delineation of the comand decision authority of the shift supervisor in the control room relative to other plant management personnel . Particular emphasis shall be placed on the following:

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when en duty in the control room. The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room.

b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to reliere the shift supervisor shall be specified.

c. If the shift supervisor is temporar.ily absent from the control room during routine operations, a lead control roca operator shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified.

3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function the shift supervisor is to provide for assuring safety. -

4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations. Admini-strative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room.

1055 32v*

h 0YT N

~

Shift Supervisor's Responsibilities (Section 2.2.la)--

Continued.

PSO COMMITMENT PS0 connits to comply with the staff position which provides methods to enhance plant safety and reliability. We recognize that the shift supervisor is the member of station management who ensures the safety and reliability of the plar.t on a daily basis. He will receive the full support of corporate management to enable him to perform his duties in a manner to provide the proper attention to safety and plant reliability.

~*"

1055 324

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Shift Technical Advisor (Section 2.2.1.b).

NRC STAFF POSITION Each licensee shall provide an on-shift technical advisor to the shift supervisor.

The shift technical advisor may serve more than one unit at a multi-unit site if qualified to perform the advisor function for the various units.

The shift technical advisor s 41' have - bachelor's degree or equivalent in a scientific or engineering disd,.line at have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shift technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.

NRC STAFF COMMITMENTS PS0 will provide an on-shift technical advisor to the on-duty shift supervisor.

The technical advisor shall have suitable experience, education and training as described in the staff position to prepare him for the duty of advising shift personnel on safe operations of the plant.

-2s_

1055 525

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Shift and Relief Turnover Procedures (Section 2.2.1.c).

NRC STAFF POSITION The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:

1. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complate and sign.

The following items, as a minimum, shall be included in tne checklist:

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist);

b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist);

c. Identification of systems and components that are in a degraded mode of operation pennitted by the Technical Specifications. For such systems and components, the length of time in degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the offgoing and oncoming auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance of test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklists);and

3. A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments).

PS0 C0MMITMENT PS0 commits to compliance with the above position and concurs that it is a prudent management approach to plant operations.

1055 326

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Control Room Access (Section 2.2.2.a).

NRC STAFF POSITION The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operationssupervisor, shift supervisor, and control room operator 3),

to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel. Provisions shall include the following:

1. Develop and implement an aaministrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access;

2. Develop and implement procedures that establish a clear line of uthority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.

PS0 COMMITMENT PS0 will comply fully with this position and recognizes the importance of access control to the control room.

1055 327

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Onsite Technical Support Center (Section 2.2.2.b).

NRC STAFF POSITION Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capabilty to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions.

The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.

A complete set of as-built drawings and other records, as described in ANSI N45.2.9-1974, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions. These Jocuments shall include, but not be limited to, general arrangement drawings, P&ID's, piping system isometrics, electrical schematics, and photographs of components installed without layout specifications (e.g., field-run piping and instrument tubing).

PS0 CCMMITMENT An onsite technical support center as described above will be provided with the capability to display necessary plant status information for individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. Various tools needed to support engineering and operational analyses shall be provided therein, such as comunications and as-built drawings. The activation and use of this center shall be governed by the BFS Emergency Plan and the plant administrative procedures.

A description of this center will be provided in the FSAR.

1055 .528

NRR Lessons Learnea Task Force Short-Term Recommendations TITLE: Onsite Operational Support Center (Section 2.2.2.c).

NRC STAFF p0SITION An area to be designated as the onsite operational support center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel will report in an emergency situation.

Communications with the control room shall be provided. The emergency plan shall be rcvised to reflect the existence of the center and to establish the methods and lines of communication and management.

pSO COMMITMENT PS0 will designate an area to serve as the operational support center as described in the above position. The support center will be physically separated from the control room, and appropriate communication facilities between the two will be pro-vided. The BFS Emergency Plan and Station administrative procedures will describe the activation and use of the Operational Support Center, as well as establish the nethods and lines of communication andranagement control. The location of the Center will be provided in the FSAR.

1055 329

NRR Lessons Learned Task Force Short-Term' Recommendations TITLE: Revised Limiting Conditions for Operation of Nuclear Power plants Based Upon Safety System Availability (Section 2.2.3 .

NRC STAFF p0SITION All NRC nuclear power plant licensees shall provide information to define a limiting operational condition based on a tnreshold of complete loss of safety function. Identification of a human or operational error that prevents or could prevent the accomplishment of a safety function required by NRC regulations and analyzed in the license application shall require placement of the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The loss of operability of a safety function shall include consideration of the ncessary instrumentation, controls, emergency electrical power sources, cooling or seal water, lubrication, operating procedures, maintenance procedures, test procedures and operator interface with the system, which must also be capable of performing their auxiliary or supporting functions. The limiting conditions for operation shall define the minimum safety functions for modes 1, 2, 3, 4, and 5 of operation.

The limiting conditions of operation shall require the following:

1. If the plant is critical, restore the safety function (if possible) and place the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />;

2. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, bring the plant to cold shutdown;

3. Determine the cause of the loss of operability of the safety function. Organizational accountability for the loss of operability of the safety system shall be established;

4. Determine corrective actions and musures to prevent recurrence of the specific loss of operability for the particular safety function and generally for any safety function;

5. Report the event within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirm by tele-graph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designee;

6. Prepare and deliver a Special Report to the NRC's Director of Nuclear Reactor Regulation and to the Director of the appropriate regional office of the Office of Inspection and Enforcement. The report shall contain the results of steps 3 and 4, above, along with a basis for allowing the plant to return to power operation. The senior corporate executive of the licensee responsible and accountable for safe plant operation shall deliver and discuss the contents of the report in a public meeting with the Office of Nuclear Reactor Regulation and the Office of Inspection and Enforcement at a location to be chosen by the Director of Nuclear Reactor Regulation.

1055 330

Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability (Section 2.2.3)--Continued.

7. A finding of adequacy of the licensee's Special Report by the Director of Nuclear Reactor Regulation will be required before the licensee returns the plant to power.

PSO COMMITMENT As indicated in the NUREG-0578 discussion preceding the position stated above, the Lessons Learned Task Force recognized that this position should be implemented through the rulemaking process provided fcr under the Administrative Procedures Act. This approach was emphasized in Dr. Mattson's letter of July 18, 1979 to Mr. Denton, attached. During the July 20 meeting with PSO, Mr. Denton stated that

~

any commitment to the position must await the rulemaking process.

In view of the foregoing, no commitment to the above position is required of PS0 at this time. PS0 does agree to comply with any requirement ultimately determined by the rulemaking.

1055 331

.. pa a%,

o, UNITED STATES y j NUCLEAR REGULATORY COMMISSION

r. WASHINGTON, D. C. 20555

• , {

p

%.....* July 18,1979 MEMORANDUM FOR: Harold R. Denton, Director Office of Haclear Reactor Regulation FROM: Roger J. Mattson, Director TMI-2 Lessons Learned Task Force

SUBJECT:

TMI-2 LESSONS LEARNED TASK FORCE REPORT (SHORT TERM) NUREG-0578 Enclosed is the first report of the TMI-2 Lessons Learned Task force.

It contains a set of short term recommendations to be implemented in two stages over the next 18 months on operating plants, plants under construction, and pending construction permit applications. There are 23 specific recomendations in 12 broad areas (nine in the area of design and analysis and three in the area of operations). The 23 recom-mendations would provide substantial, additional protection which is required for the public health and safety.

All but one of the 23 recommendations have a majority concurrence by the Task Force. The exception is the recommended requirement to provide capability to install an external recombiner at each reactor plant forThe post-accident hydrogen control, if necessary following an accident.

majority of the Task Force recommends that this matter daserves further evaluation in conjunction with other hydrogen generation and control questions being reviewed by the Task Force for its final report.

Three of the recommendations appear to require changes in existing regulations for which the Task Force recomends immediately effective rulemaking. They are: 1) inerting of MKJ and MK II BWR containments that are not already inerted; 2) provision of the capability to install an external recombiner for plants that do not already have recombiners (minority view); and, 3) revised limiting conditions of operation in operating licenses for total loss of safety system availability through The Office of Standards Development has agreed human or operational error.

to develop the requ' red Commission papers and carry through with these rulemaking actionr The 23 recommended actions were discussed with the Regulatory Requirements Review Committee (June 22, 1979), the Commission (June 25,1979),the THI-2 Subcommittee of the ACRS (July 11,1979), and the ACRS (July 12, 1979).

In addition, meetings were held with various groups in the Office of Nuclear Reactor Regulation in the course of the last few weeks to discuss technical aspects of specific portions of the recommended actions and the implementation alternatives.

1055 532 77D75/co6

2 Harold R. Dcnton The Task Force recommends that time not be taken to request and evaluate public comments on these short term requirements prior to their promulgation as licensing requirements or rules because they are safety significant matters that require prompt application to operatingOther reactors and THI-2 operating accident license review grou'ps applications in the late stages of review.and the Lessons Learned Task implications of the accident. Any public coments on the short term recom-mendations that are received after their issuance (just as in the case of the earlier IE Bulletins) can be factored into those continuing evaluations.

Having identified the 23 specific recomendations for short term action, the Lessons Learned Task Force will turn to the broader, more fundamental regulatory questions which should be addressed in the longer term (some of them likely to require evaluations that extend beyond the life These span of the longer Task Force) before other regulatory actions are recommended. -

term interests The TaskofForce the Task intendsForce are its to develop described in Section Three final recommendations and of the report. The topics to be addressed issue a final report in early September 1979.

in the final report could affect the future structure and content of the licensing process to correct deficiencies identified by the TMI-2 accident and to further upgrade the level of safety in operating plants and plants The Task Force does not believe that allowing new plants under construction.

to begin operation in the next few months will foreclose further design changes that may be shown to be desirable by its continuing review of the accident.

On July ll, I solicited the comments of the principal NRR line organizations on the final draft of the report and its central conclusion regarding the necessity and sufficiency of the short term recommendations for continued operations and licensing. General support for the conclusions of the Task Force report was expressed by all of the principal NRR line managers.

We have reviewed and considered the detailed comments supplied Wherebyappropriate, the various NRR organizations in the course of their review. The principal we made clarifying changes in the language of the report.

substantive change occurred in the form and schedules of the implementaticn section (Appendix B). Some of the comments addressed matters There arethat the Task Force has deferred for consideration in its final report.

significant differences of opinion within the staff on two of the Task recommendations, as follows:

rulemaking for revised limiting conditions for operation (some agree with the recomendation and others prefer more stringent enforcement actions using existing regulatory machinery) ard b) the need for the minority Task Force recommendation 2.1.5.c concerning rulemaking for backfit of recombiner capability (some support the minority recomendation, others do Having considered these comments and made changes to the report where not).

appropriate to reconcile them with the intent of the Task Force, I recomend that you:

a. direct the immediate implementation by DPM, 00R or B&OTF, as appropriate, of all the short term recommendations, except the tnree rulemak matters, through the issuance of licensing positions to operating plant licensees, plants under construction, and construction permit applicants.

1055 533

3 Harold R. Denton

b. request the fomulation of imediately effective rules by the Office of Standards Development for action by the Commission on the three rulemaking matters.

Another matter that needs to be considered by you in deciding upon the additional requirements for near term CP and OL decisions and for operating reactors is improvements in licensee emergency preparedness.

p Roger J.

tattson, Director TMI-2 Lessons Learned Task Force .

Enclosure:

as stated cc: Chaiman Hendrie

• Commissioner Gilinsky Commissioner Kennedy '

Commissioner Bradford Commissioner Ahearne ACRS (20)

Policy Evaluation SECY L. V. Gossick, ED0 S. Levine, RES R. Minogue, SD V. Stello, IE M. Rogovin, Special Inquiry J. Fouchard, PA (20)

C. Kammerer, CA (20)

NRC PDR .

g

e 9 e

RESPONSE TO INSPECTION & ENFORCEMENT BULLETIN 79-08, c

e t-J ,,e l U, D s3D

IEB 79-08 Task 1 Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 03/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.

a. This review should be directed toward understanding: (1) the extreme seriousness and consequences of the simultaneous blocking of both trains of a safety system at the Three Mile Island Unit 2 plant and other actions taken during the early. phases of the accident; (2) the apparent operational errors which led to eventual core damage; and (3) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action;

b. Operational personnel should be instru:ted to: (1) not override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a of this bulletin); and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available;

c. All licensed operators and Plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be docunentedin plant records.

PSO COMMITMENT Public Service Company of Oklahoma has established a Technical Advisory Committee (TAC) to assess the events at Three Mile Island, Unit 2, and to apply the lessons learned to its Black Fox Station Project. This contaittee was established at the direction of the President and Chief Executive Officer of the Company and reports its findings and recommendations directly to the Review and Audit Committee.

These findings and recommendations will then be implemented by the Review and Audit Comittee.

The TAC has been directed to utilize PS0 and consultant resources to fully review the interim and final results of the various investigations. These presently include:

. USNRC's " Lessons Learned Task Force"--NUREG-0578

. The President's Commission on Three Mile Island EPRI--Nuclear Safety Analysis Center . ,

}ts J _. 3 0 IEB 79-08 Task 1--Continued.

. Generic vendor programs

. Atomic Industrial Forum TMI Policy Committee

. NRC Special Invest.igation (Rogovin)

The TAC and its consultants have already assessed issuances of the ACRS and regulatory staff and presented a preliminary assessment to the NRC Staff in our June 15 submittal. It is aware of the activities of various other legislative and regulatory investigations and will assess future recomendations from them.

The assessmenc and resulting program was predicated on the advice, and guidance set forth in the various letters, from the ACRS (particularly their letters of April 7 and May 16, 1979), and IE Bulletin No. 79-08, dated April 14, 1979. In addition, S. Levy, Inc., a participant in both the post-event safe shutdown activities of TMI and the EPRI investigation, has been retained to keep PS0 continously informed of any new developments arising from the ongoing investigations by EPRI and other organizations.

The objective of the TAC and its consultants is to ensure that the Black Fox Sation design, construction, operating procedures, staffing and training program, and emergency response plan incorporates the benefits of the TMI investigation to the fullest extent practicable.

The effort is directed toward understanding: (1) the extreme seriousness and consequences of the simultaneous blocking of both trains of a safety system at the Three-Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to eventual core damage; and (3) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action.

Prior to completion of operating procedures and training instructions for j g . r) ]7 operation of the Black Fox Station, these procedures and instructions will be reviewed to assure that operational personnel are instructed to: (1) not override

IEB 79-08 Task 1--Continued. automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions, and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are availcble. (See also commitments made under IEB 79-08 Task 5). The Manager, Black Fox Station and the -Manager, Nuclear Training are assigned to the TAC to ensure that operational experience is considered in the TAC reviews and to provide continuity for implementation of TAC findings into operator license and station supervisor / management training. A key objective of the TAC is to review administrative mechanisms to ensure that lessons learned are incorporated into the station training programs. Findings and reconmendations from the TAC will be documented in the Project files and confomance with each specified commitment will be incorporated into this documentation system. 1055 338

IEB 79-08 Task 2 Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate containment isolation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection. PS0 COMMITMENT At the time of final design, i.e., FSAR submittal, and prior to completion of operating procedures, containment isolation initiation will be reviewed to assure containment isolation of all lines whose isolation does not degrade needed safety features or cooling capability upon automatic initiation of safety injection. This isolation may be automatic or manual, and any necessary manual actions will be covered by appropriate procedures. 1055 339

IEB 79-08 Task 3 Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e.g., RCIC) that are used when the main feedwater system is not operable. For any manual action necessary, describe in summary form the procedure by which this action is taken in a timely sense. PSO COMMITMENT At the time of final design, i.e, FSAR' submittal, and prior to completion of operating procedures, the functioning of the auxiliary heat removal systens that are used when the main feedwater system is not operable will be reviewed. Both automatic and manual actions will be assessed for adequacy, and any necessary manual actions will be addressed by procadures to assure timely actuations. 1055 540

IEB 79-08 Task 4 Describe all uses and types oT vessel level indicatinn for both automatic and manual initiation of safety systems. Describe other redundant instrumentation which the operator might have to give the same infonnation regarding plant status. Instruct operators to utilize other available information to initiate safety systems. pSO COMMITMENT At the time of final design, i.e, FSAR submittal, and prior to completion of operating procedures, all uses and types of vessel level indication for both Redundant automatic and manual initiation of safety systems will be reviewed. instrumentation which the operator will have to give the same vessel level indications will be identified and factored into operator training, instruction, and procedures. 10 5'i 3 /i l

IEB 79-08 Task 5 Review the action directed by the operating procedures and training instructions to ensure that:

a. Operators do not override automatic actions of engineered safety features, unless continued operation of engineered safety features will result in unsafe plant conditions (e.g., vessel integrity);

b. Operators are provided additional information and instructions to not rely upon vessel level indicat. ion alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.

PSO COMMITMENT Prior to completion of operating procedures and training instructions, actions directed by these instructions will be reviewed to ensure that:

a. Operators are directed not to override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions;

b. Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.

1055 !,42

IEB 79-08 Task 6 Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also, review related procedures, such as those for maintenance, testing, plant and system startup, and supervisory periodic (e.g., daily / shift checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes. pSO COMMITMENT At the time of final design, i.e., FSAR submittal, PS0 will review all safety-related valve positioning requirements and positive controls to assure that valves remain positioned in a manner to ensure the proper operation of engineered safety features. In addition, prior to completion of related procedures, the precedures for maintenance, testing, plant and systems startup, and supervisory periodic surveillance will be reviewed to ensure that safety-related valves are returned to the correct position following necessary manipulations and are maintained in the proper position during all operational modes.

                                                                                                                         }0;) sk3

IEB 79-08 Task 7 Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary contain-ment to assure that undesired pumping, venting, or other relase of radioactive liquids and gases will not occur inadvertently. In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate:

a. Whether interlocks exist to prevent transfer when high radiation indication

                                        ~

exists, and;

b. Whether such systems are isolated by the containment isolation signal;

c. The basis on which continued operability of the above features is a' lured.

PS0 COMMITMENT At the time of final design, i.e., FSAR submittal, and prior to completion of operating procedures, the operating modes of all systems designed to transfer potnetially radioactive gases and liquids out of the primary containment will be reviewed to assure that undesired pumping, venting, or otner release of radioactive gases and liquids will not occur inadvertently. In particular, the impact of resetting of engineered safety features instrumentation will be examined to ensure that such an inadvertent radioactive liquid or gas release will not result from this resetting. Each of the above systems will be reviewed to assure that:

a. Interlocks exist to prevent transfer when high radiation indication exists, and;

b. Such systems are isolated by the containment isolation signal.

                                          -4 0-

IEB 79-08 Task 8 Review and modify as necessary your maintenance and test procedures to ensure that they require:

a. Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service;

b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing:

c. Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.

PS0 COMMITMENT Prior to their completion, maintenance and test procedures for safety-related systems will be reviewed to ensure that they require:

a. Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service;

b. Verification of the operability of all safety-related syste is when they are returned to service following maintenance or testing;

c. Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from or returned to service.

1055 345

IEB 79-08 Task 9 Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation. Further, at that time, an open continuous comunication channel shall be established and maintained with NRC. PS0 COMMITMENT l and implementing procedures, NRC Prior to completion of the emergency p' an notification shall be incorporated to assure that NRC is notiff.ed within one hour of the time the reactor is not in a controlled or expected condition of operation. Further, at the time of NRC notification, an open continuous comunication channel will be established and maintained with NRC.

                                          -42                      1055 346

IEB 79-08 Task 10 Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment. PS0 COMMITMENT At the tine of final design, i.e, FSAR submittal, and prior to completion of operating procedures, operating modes and procedures will be reviewed to assure that they are adequate to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the . primary" system or be released to the containment. 1055 547

IEB 79-08 Task 11 Propose changes, as required, to those technical pecifications which must be modified as a result of your implementing the items above. PS0 COMMITMENT Those issues that need to be addressed by technical specifications as a result of implementing IEB 79-08 task items 1 through 10 shall be incorporated prior to completion of the technical specifications which will be submitted with the FSAR.

                                            #~

1055 348

RESPONSE TO SELECTED ISSUES ON EMERGENCY PREPAREDNESS lnC; us, .'; k )

Emercency Preparedness

1. Regulatory Guide 1.101 Emergency Plannino For Nuclear Power Plants.

The BFS PSAR, Section 1.9 reflects a commitment to revision 0 of this regulatory guide. For the purposes of design and development of operating procedures, PS0 will use Revision 1 dated March, 1977. Full implementation will be demonstrated at the time of FSAR submittal. Discussions with the regulatorv staff have indicated that revisions to the uniform action level criteria will be forthcoming as a result of the experiences at TW . PS0 will utilize these criteria in development of the BFS Emergency Plan. ii. Improved Sampling and Instrumentation Capability. These issues are covered in NUREG-0578 TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations as issue 2.1.8. PS0 has addressed these requirements in our response to that section. iii. Emeraency Operating Center. The BFS PSAR % 13.3.3 identifies a secondary Emergency Control Center located away from the generation complex, but within the site boundary. This center will serve as the focal point for radiological emergency response, i.e., an emergency operating center, by being the coordination point for local, state, and federal authorities involved. Appropriate plant status and meteorological data will be read directly from instrumentation placed in the center, iv. Improved Offsite Monitoring capability. As a part of its evaluation of the events at TMI, PS0 comits to reevaluate the necessary capabilities of offsite radiation monitors. The number and location of thermoluminescent dosimeters (TLD's) will be studied, as well as _4s_ 1055 350

Emergency Perparedness - iv. (Continued). possible use of continuous radiation monitors with remote readout. PS0 also comits to closely monitor forthcoming regulatory guidance in this area to assure that appropriate capabilities are promptly factored into the BFS design and operation plan.

v. Adequacy of Protective Action Planning.

PS0 is evaluating the current regualtory requirements for emergency planning in light of the events at TMI. Since April 1,1979, our techincal staff has had several meetings with Oklahoma State Department of Health, Division of Occupational and Radiological Safety personnel who have been designated by the Governor, State of Oklahoma, as the prime state agency respondent. The State of Oklahoma does not presently have in effect an emergency response plan. The attached letter dated June 20, 1979 from George Nigh, Governor, State of Oklahoma, to Joseph Hendrie, Chairman, U. S. Nuclear Regulatory Co mission, explains the State's status in preparing such a plan, and receiving NRC approval. As stated therein, PS0 personnel are working closely with the State in review of the draft. We are fully prepared to assist the State in timely final development

   .nd submittal to NRC approval.

Concurrently, P50 is establishing target tasks for the BFS Emergency Response Plan development. The plan will be submitted with the FSAR in support of the application for operating licenses. Our understanding from recent discussior.s with the Staff is that protective actions in the future may be planned out to a radius of 10 miles rather than out to the radius of the Low Population Zone (LPZ) of 4,000 meters as reflected in the BFS Preliminary Safety Analysis Report and Environmental Report. Accordingly, we have reviewed the applicable discussion from the ER ( 2.1.3.1) on the popluation prcjections within a ten-mile radius of the site. Also studied were PSAR tabluations of regional incorporated community statistics and population 1055 351

Emergency Preparedness - v. (Continued). projections of the two communities within the area. Finally, we examined the PSAR figure relating to emergency evacuation routes for the ten-mile area. The only significant population concentration within the ten-mile radius area is the town of Inola. The area is primarily rural and is expected to remain so during the lifetime of Black Fox Station. The 1980 estimated population of Inola is 2900 with projections increasing to 4600 by the year 2020. There are three other small communities within ten miles of Black Fox Station, in addition to Inola as shown in ER figure 2-1-6. They are New Tulsa (eight miles WSW), Fair Oaks (nine miles WNW), and Tiawah (ten miles N). New Tulsa and Fair Oaks populations are expected to increase only marginally. Much of the Tiawah 1980 estimated population of 125 is located beyond the ten-mile radius while the 2020 population is expected to be only 321. The accompanying ER Table 2-1-1 shows that the overall population density within the tan-mile radius of the Black Fox Station is small--less than 15,000 in 1980 and less than.24,000 in 2020. pSAR Figure 13.3-3 shows the potential emergency evacuation routes. Major routes such as state highways 18 and 33 and U. S. Highway 69 are identified. In addition, since Oklahoma is unifomly divided into square mile sections, each of the perpendicular lines foming uniform squares on the figure represents a transportation route. As a result of our review, we have concluded that implementation of protective measures such'as evacuation is feasible over the lifetime of the station based on population estimates and evacuation routes. vi. periodic Testing. PS0 connents to periodically conduct local emergency plan testing to assure that 1055 352

Emergency Preparedness - iv. (Continued). the plan is fully functional and kept up-to-date with regard to local population location and transportation routes. In addition, we recognize the benefits of an integrated PSO/ State /NRC test to fully check comunications and to insure correct agency interaction. We will support the practice of integrated testing.

                                           -4 3-1055 353}}