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{{#Wiki_filter:'w-s 3 6212 DIN 8-016-854 PUBUC S%1CE COMPANY OF OKLAHOMA                                           I V
{{#Wiki_filter:'w-s 3
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6212 DIN 8-016-854 PUBUC S%1CE COMPANY OF OKLAHOMA V '}- q I
                                                                                          '}- q A CENTRAL AND SOUTH WEST COMPANY                                         ;a               ,
A CENTRAL AND SOUTH WEST COMPANY f
s P.O. box 201/ TULSA. OKLAHOMA 74102 / (918) 583-3811 L           c Public Service Company of Oklahoma                                   July 27, 1979 Black Fox Station                                                     File: 6212.125.3500.21L Response to lessons Learned Report                                           6212.217.0521.21L USNRC Docket Nos. STN 50-556, 50-557 Mr. Steven A. Varga, Assistant Director Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.               20002
;a P.O. box 201/ TULSA. OKLAHOMA 74102 / (918) 583-3811 L
s c
Public Service Company of Oklahoma July 27, 1979 Black Fox Station File: 6212.125.3500.21L Response to lessons Learned Report 6212.217.0521.21L USNRC Docket Nos. STN 50-556, 50-557 Mr. Steven A. Varga, Assistant Director Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20002


==Dear Mr. Varga:==
==Dear Mr. Varga:==
 
The meeting between our respective organizations on July 19 in Bethesda was very productive. We especially appreciated your positive comments concerning our analysis of the lessons to be learned from TMI-2 as they apply to the Construction Permit application for the Black Fox Station. This analysis, which was sent to Mr. Harold R.
The meeting between our respective organizations on July 19 in Bethesda was very               -
Denton, Director, Nuclear Reaction Regulation, on June 15, 1979, represented the initial effort by Public Service Company of Oklahoma (PS0) to respond to the events at TMI and documented our long-term corporate commitment to fully analyze every facet of the TMI-2 accident arid to incorporate the lessons learned into the design, construction, staffing, training and operation of the Black Fox Station.
productive. We especially appreciated your positive comments concerning our analysis of the lessons to be learned from TMI-2 as they apply to the Construction Permit application for the Black Fox Station. This analysis, which was sent to Mr. Harold R. Denton, Director, Nuclear Reaction Regulation, on June 15, 1979, represented the initial effort by Public Service Company of Oklahoma (PS0) to respond to the events at TMI and documented our long-term corporate commitment to fully analyze every facet of the TMI-2 accident arid to incorporate the lessons learned into the design, construction, staffing, training and operation of the Black Fox Station.
With the issuance on July 19 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," e compared the 23 lessons learned in that report with the PSO analysis. Our understanding of NUREG-0578 and the comparison of the two documents were greatly facilitated by the helpful explanations and advice offered by you and your staff during our meeting on July 19. These discussions, along with inforration provided by Mr. Denton during his meeting with our President, Mr. R. O. Newman on July 20, enable us to respond promptly to your request for com-mitments to the requirements and recommendations of NUREG-0578.
With the issuance on July 19 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," e compared the 23 lessons learned in that report with the PSO analysis. Our understanding of NUREG-0578 and the comparison of the two documents were greatly facilitated by the helpful explanations and advice offered by you and your staff during our meeting on July 19. These discussions, along with inforration provided by Mr. Denton during his meeting with our President, Mr. R. O. Newman on July 20, enable us to respond promptly to your request for com-mitments to the requirements and recommendations of NUREG-0578.
Although our June 15 lessons learned analysis adcrassed most of the issues discussed in NUREG-0578, the organization of the material is different. Consequently, to facilitate your review, we are reiterating our concitments in a format consistent with the organization of NUREG-0578. In addition to specifically addressing every recommendation and requirement of NUREG-0578, this submittal also addresses matters applicable to Black Fox which were developed by the Bulletins and Orders Task Force, and the Emergency Preparedness group headed by Mr. Brian Grimes.
Although our June 15 lessons learned analysis adcrassed most of the issues discussed in NUREG-0578, the organization of the material is different. Consequently, to facilitate your review, we are reiterating our concitments in a format consistent with the organization of NUREG-0578.
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In addition to specifically addressing every recommendation and requirement of NUREG-0578, this submittal also addresses matters applicable to Black Fox which were developed by the Bulletins and Orders Task Force, and the Emergency Preparedness group headed by Mr. Brian Grimes.
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a .
a Mr. Steven A. Varga, Assistant Director Page 2.
Mr. Steven A. Varga, Assistant Director                                 Page 2.
We concur with the view presented during the meetings of July 19 and 20, that all of the commitments and actions raquired of us 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 for the Black Fox Station. Our commitments reflect this understanding and philosophy.
We concur with the view presented during the meetings of July 19 and 20, that all of the commitments and actions raquired of us 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 for the Black Fox Station. Our commitments reflect this understanding and philosophy.
The TMI-2 accident has stalled progress on the Black Fox application, and as you know, we are quite anxious to overcome this licensing delay. Consequently, we have responded directly and completely to all of the issues applicable to the Black Fox Station as presented by the two Task Forces and Mr. Grimes's group; this submittal should satisfy all of those concerns. In these circumstances, we do believe it reasonable to expect the NRC Staff to complete its report quickly and to respond to the Licensing Board Orderof June 13, 1979 in the very near future.
The TMI-2 accident has stalled progress on the Black Fox application, and as you know, we are quite anxious to overcome this licensing delay.
Consequently, we have responded directly and completely to all of the issues applicable to the Black Fox Station as presented by the two Task Forces and Mr. Grimes's group; this submittal should satisfy all of those concerns.
In these circumstances, we do believe it reasonable to expect the NRC Staff to complete its report quickly and to respond to the Licensing Board Orderof June 13, 1979 in the very near future.
Please call Mr. Vaughn Conrad, Manager, Licensing and Compliance at (918) 583-3611 if you have any questions regarding this submittal.
Please call Mr. Vaughn Conrad, Manager, Licensing and Compliance at (918) 583-3611 if you have any questions regarding this submittal.
Sincerely yours, b[ _.
Sincerely yours, b[ _.
T. N. Ewing, Manager Black Fox Stationplear Project TNE:VLC:dm Attachment xc: (w/ attachment) BFS Service List (o '{
T. N. Ewing, Manager Black Fox Stationplear Project TNE:VLC:dm Attachment xc:
(w/ attachment) BFS Service List (o '{


BLACK FOX STATION SERVICE LIST CERTIFICATE OF SERVICE I hereby certify that a copy of the foregoing PS0 Response to the TMI Event has been served on each of the following persons by deposit in the United States mail, first-class postage prepaid, this 27th day of July, 1979.
BLACK FOX STATION SERVICE LIST CERTIFICATE OF SERVICE I hereby certify that a copy of the foregoing PS0 Response to the TMI Event has been served on each of the following persons by deposit in the United States mail, first-class postage prepaid, this 27th day of July, 1979.
L. Dow Davis, Esquire                                 Mr. Joseph Gallo Counsel for NRC Staff                                 Isham, Lincoln & Beale U. S. Nuclear Regulatory Comission                   105017th Street N. W.
L. Dow Davis, Esquire Mr. Joseph Gallo Counsel for NRC Staff Isham, Lincoln & Beale U. S. Nuclear Regulatory Comission 105017th Street N. W.
Washington, D. C.       20036 Washington, D. C.       20555 Mr. Cecil 0. Thomas                                   Joseph R. Farris, Esquire U. S. Nuclear Regulatory Cormiission                 Green, Feldman, Hall & Woodard Phillips Building                                     816 Enterprise Building 7920 Norfolk Avenue                                   Tulsa, Oklahoma       74103 Bethesda, Maryland         20014 Docketing and Service Section                         Andrew T. Dalton, Esquire Office of the Secretary of the Comn.                 1437 South Rin Street, Suite 302 U. S. Nuclear Regulatory Comission                   Tulsa, Oklahoma       74119 Washington, D. C.           20555 (20 copies)                                                     -
Washington, D. C.
Mr. William G. Hubacek                               Mrs. Ilene H. Youngnein U. S. Nuclear Regulatory Commission                   3900 Cashion Place Office of Inspection and Enforcement                   Oklahoma City, Oklahoma       73112 Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas         76012 Mr. Gerald F. Diddle                                 Mr. Lawrence Burrell General Manager                                       Route L, Box 197 Associated Electric Cooperative, Inc.                 Fairview, Oklahoma       73737 P. 0. Box 754 Springfield, Missouri         65801 Mr. Maynard Human                                     Mrs. Carrie Dickerson General Manager                                       Citizens' Action for Safe Energy, Inc.
20555 Washington, D. C.
Western Famers Electric Ccoperative                   P. O. Box 924 P. O. Box 429                                         Claremore, Oklahoma       74017 Anadarko, Oklahoma         73005 Michael I. Miller, Esq.                               Charles S. Rogers, Esq.
20036 Mr. Cecil 0. Thomas Joseph R. Farris, Esquire U. S. Nuclear Regulatory Cormiission Green, Feldman, Hall & Woodard Phillips Building 816 Enterprise Building 7920 Norfolk Avenue Tulsa, Oklahoma 74103 Bethesda, Maryland 20014 Docketing and Service Section Andrew T. Dalton, Esquire Office of the Secretary of the Comn.
Isham, Lincoln & Beale                               Assistant Attorney General Cne 1st National Plaza                                 112 State Capitol Building Suite 4200                                             Oklahoma City, Oklat ma , 7 105     g Chicago, Illinois       60603                           g                         )       i OA%%rs iW '
1437 South Rin Street, Suite 302 U. S. Nuclear Regulatory Comission Tulsa, Oklahoma 74119 Washington, D. C.
20555 (20 copies)
Mr. William G. Hubacek Mrs. Ilene H. Youngnein U. S. Nuclear Regulatory Commission 3900 Cashion Place Office of Inspection and Enforcement Oklahoma City, Oklahoma 73112 Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76012 Mr. Gerald F. Diddle Mr. Lawrence Burrell General Manager Route L, Box 197 Associated Electric Cooperative, Inc.
Fairview, Oklahoma 73737 P. 0. Box 754 Springfield, Missouri 65801 Mr. Maynard Human Mrs. Carrie Dickerson General Manager Citizens' Action for Safe Energy, Inc.
Western Famers Electric Ccoperative P. O. Box 924 P. O. Box 429 Claremore, Oklahoma 74017 Anadarko, Oklahoma 73005 Michael I. Miller, Esq.
Charles S. Rogers, Esq.
Isham, Lincoln & Beale Assistant Attorney General Cne 1st National Plaza 112 State Capitol Building Suite 4200 Oklahoma City, Oklat ma, 7 105 g
Chicago, Illinois 60603 g
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i OA%%rs iW Manager,LicensinN(\\
Vaughn L.qonr3t!
Vaughn L.qonr3t!
Manager,LicensinN(\  nd Compliance
nd Compliance
[g 5                       Public Service Company of Oklahoma
[g 5 Public Service Company of Oklahoma


RESPONSE OF PUBLIC SERVICE COMPANY OF OKLAHOMA BLACK FOX STATI0f:, UNITS 1 & 2 USNRC DOCKET N05. 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 Issues on Emergency Preparedness July 27, 1979.
RESPONSE OF PUBLIC SERVICE COMPANY OF OKLAHOMA BLACK FOX STATI0f:, UNITS 1 & 2 USNRC DOCKET N05. 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 Issues on Emergency Preparedness July 27, 1979.
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TABLE OF CONTENTS                                 Page Introduction & Description of Methodology. . . . . . . . . . . . . . . . . . .       1 Response to NUREG-0578, Appendix A. . . . . . . . . . . . . . . . . . .        ...
TABLE OF CONTENTS Page Introduction & Description of Methodology...................
Section 2.1.1     Emergency Power Supply Requirements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves and Pressurizer Level Indicators in PWR's. . . . . . . . . . . . . . . . . . 4 2.1.2     Performance Testing for BWR and PWR Relief and Safety Va l v es . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.1.3.a   Direct Indication of Power-0perated Relief Valves and Safet3 Valve Position for PWR's and BWR's. . . . . . . . . . . . .       6 2.1.3.b   Instrumntation for Detection of Inadequate Core Cooling in PWR's and BWR's. . . . . . . . . . . . . . . . . . . . . . . 7 2.1.4     Diverse and More Selective Containment Isolation Prcvisions for PWR's and BWR's. . . . . . . . . . . . . . . . . . . . . . . 9 2.1.5.a   Dedicated Penetrations for External Recombiners or Post-Accident P u rg e Sys tems . . . . . . . . . . . . . . . . . . . . . . . . 10 2.1.5.b   Inerti ng BWR Containments. . . . . . . . . . . . . . . . . . . . 11 2.1.5.c   Capability to Install Hydrcgen Recombiner at Each Light Water Nuclear Power Plant. . . . . . . . . . . . . . . . . . . . . 12 2.1.6.a   Integrity of Systems Ottside 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.s.b   Auxiliary Feedwater Flow Indication to Steam Generators for PWR's. . . . . . . . . . . . . . . . . . . . . . . . . . . .       16 1.1.8.a   Improved Post-Accident Sampling Capability. . . . . . . . . . .         17 2.1.8.3   Increased Range of Radiation Monitors. . . . . . . . . . . . .         .18
1 Response to NUREG-0578, Appendix A...................
Section 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves and Pressurizer Level Indicators in PWR's.................. 4 2.1.2 Performance Testing for BWR and PWR Relief and Safety Va l v es...........................
5 2.1.3.a Direct Indication of Power-0perated Relief Valves and Safet3 Valve Position for PWR's and BWR's.............
6 2.1.3.b Instrumntation for Detection of Inadequate Core Cooling in PWR's and BWR's....................... 7 2.1.4 Diverse and More Selective Containment Isolation Prcvisions for PWR's and BWR's....................... 9 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident P u rg e Sys tems........................ 10 2.1.5.b Inerti ng BWR Containments.................... 11 2.1.5.c Capability to Install Hydrcgen Recombiner at Each Light Water Nuclear Power Plant..................... 12 2.1.6.a Integrity of Systems Ottside 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.s.b Auxiliary Feedwater Flow Indication to Steam Generators for PWR's............................
16 1.1.8.a Improved Post-Accident Sampling Capability...........
17 2.1.8.3 Increased Range of Radiation Monitors.............
.18


Table of Contents                                                           Page 2.1.8.c   Improved In-Plant Iodine Instrumentation. . . . . . . . . . . 19 2.1.9     Analysis of Design and Off-Normal Transients and 20 Accidents. . . . . . . . . . . . . . . . . . . . . . . .
Table of Contents Page 2.1.8.c Improved In-Plant Iodine Instrumentation...........
2.2.1.a   Shift Supervisor's Responsibilities. . . . . . . . . . . . . 23 2.2.1 b   Shift Technical Advisor. . . . . . . . . . . . . . . . . . . 25 2.2.1.c   Shift and Relief Turruver 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 Safaty Syrtem nvailability. . . . . . . 30 Response to Inspection and Enforcem:nt Bulletin 79-08. . . . . . . . . . . 32 Response to Selected Issucs on Emergency Preparedness. . . . . . . . . . . 45 0   >
19 2.1.9 Analysis of Design and Off-Normal Transients and Accidents........................
20 2.2.1.a Shift Supervisor's Responsibilities.............
23 2.2.1 b Shift Technical Advisor...................
25 2.2.1.c Shift and Relief Turruver 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 Safaty Syrtem nvailability.......
30 Response to Inspection and Enforcem:nt Bulletin 79-08...........
32 Response to Selected Issucs on Emergency Preparedness...........
45 0


INTRODUCTION AND DESCRIPTION OF METHODOLOGY On June 15, 1979, PuMic 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 tL events at TMI and to implement the lessons learned into our project.
INTRODUCTION AND DESCRIPTION OF METHODOLOGY On June 15, 1979, PuMic 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.
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.
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 tL 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 Committee 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.
Prior to development of this document, consultants to and members of the Technical Advisory Committee 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.
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, h,
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, h,
For the balance, the intent of each commitment by P50 is to meet the express position of the regulatory staff as stated in NUREG-0578, Appendix A.
For the balance, the intent of each commitment by P50 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 neeting 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 Bulletiri (IEB) 79-08.
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.
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 commitment for BFS.
In the neeting 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 Bulletiri (IEB) 79-08.
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 cperation. The PS0 commitnents to action require corpletion of the efforts described during final design as detailed in the FSAR and in subsequently developed operating procedures.
The June 15 submittal by PS0 was intended to incorporate all of the requirements statedin IEB 79-08.
PS0 recognizes that the " Lessons Learned" requirements and the IEB 79-08 requirements represent separate activities within the regulatory staff. Thus, there exists scme duplication of subject matter with the possibility of different interpretations of the PSO response between the two task forces.     If such differences are identified, PSO commits to work with the NRC Staff to reconcile them.
In order to be completely responsive, each of the IEB 79-08 Tasks are repeated in this submittal followed by the appropriate PS0 commitment for BFS.
                                  ~
The IEB 79-08 was specifically addressed to licensees with operating boiling water reactors and response was required very quickly.
                              -;L'
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 cperation.
The PS0 commitnents to action require corpletion 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 scme duplication of subject matter with the possibility of different interpretations of the PSO response between the two task forces.
If such differences are identified, PSO commits to work with the NRC Staff to reconcile them.
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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 PS0 should address in this submittal. Most were covered in our June 15 assessment.
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 PS0 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 Therein, of the State of Oklahoma, George Nigh to Joseph Hendrie, Chairman USNRC.
Included in the emergency preparedness section is a letter from the Governor
: Therein, of the State of Oklahoma, George Nigh to Joseph Hendrie, Chairman USNRC.
the status of the State Emergency Response Plan, PS0's role in development, and a connitment to have a NRC approved plan in effect well before BFS conmercial operation is discussed.
the status of the State Emergency Response Plan, PS0's role in development, and a connitment to have a NRC approved plan in effect well before BFS conmercial 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.
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.
Line 84: Line 130:


NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Performance Testina for BWR and PWR Relief and Safety Valves (Section 2.1.2).
NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Performance Testina 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 maximized. 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.
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 maximized. 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 d
PS0 COMMITMENT d
PS0 believes that it is important to assure that the safety and relief valves installed in the BFS reactor coolant bnundary will function as intended and maintain their integrity under exoected 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 necesssry testing to qualify the reactor coolant system relief and safety valves for the ecst severe conditions identified.
PS0 believes that it is important to assure that the safety and relief valves installed in the BFS reactor coolant bnundary will function as intended and maintain their integrity under exoected 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 necesssry testing to qualify the reactor coolant system relief and safety valves for the ecst 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.
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 applicaticn.
Documentation will be contained in the FSAR at the time of submittal in support of the operating license applicaticn.
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Direct Indication of Power-Ocerated Relief Valve and Safety Valve Position for PWR's and BWR's Section 2.1.3.a   .
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Direct Indication of Power-Ocerated Relief Valve and Safety Valve Position for PWR's and BWR's Section 2.1.3.a NRC STAFF POSITION Reactor system relief and safety valves shall be provided with a positive indi-cation in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe.
NRC STAFF POSITION Reactor system relief and safety valves shall be provided with a positive indi-cation in the control room derived from a reliable valve position detection 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.
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 F5AR.
Design detail will be provided in the F5AR. 75
75


NRR Lessons Learned Task Force Short-Term Reco mendations TITLE:   Instrumentation for Detection of Inadeouate Core Cooling in PWR's and BWR's (Section 2.1.3.b   .
NRR Lessons Learned Task Force Short-Term Reco mendations TITLE:
NRC STAFF POSITION
Instrumentation for Detection of Inadeouate Core Cooling in PWR's and BWR's (Section 2.1.3.b NRC STAFF POSITION 1.
: 1. Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation.
Licensees shall develop procedures 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 descriptirn of the analyses needed to form the basis for operator training and procedure development shall be provided ,1ursuant to another short-term requirement, " Analysis of Off-Nomal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).
The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions. A detailed descriptirn of the analyses needed to form the basis for operator training and procedure development shall be provided,1ursuant to another short-term requirement, " Analysis of Off-Nomal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).
In addition, each PWR shall install a primary coolant saturation meter 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.
In addition, each PWR shall install a primary coolant saturation meter 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       .
2.
those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included.
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 of the functional design requirements for the system shall also be included.
A descripticn 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.
A descripticn 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 unambiguously deternine 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 cperator in recognizing inadequata core cooling, and oparators will be throroughly trained in the procedure and utilization of instrumentation to assure correct interpretation of the core h
PS0 COMMITMENT The ability of station operators to easily and unambiguously deternine 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 cperator in recognizing inadequata core cooling, and oparators will be throroughly trained in the procedure and utilization of instrumentation to assure correct interpretation of the core h
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.
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.
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Containment Isolation Provisions for PWR's and BWR's (Section 2.1.4).
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
NRC STAFF POSITION 1.
: 1. All containment isolation system designs shall comply with the recontendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.
All containment isolation system designs shall comply with the recontendations 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 accordingly, and shall report the results of the re-evaluation to the NRC.
2.
: 3. All non-essential systems shall be automatically isolated by the containment isolation signal.
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 accordingly, and shall report the results of the re-evaluation to the NRC.
: 4. The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of con-tainmentisolation valves shall require deliberate operator action.
3.
All non-essential systems shall be automatically isolated by the containment isolation signal.
4.
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of con-tainmentisolation valves shall require deliberate operator action.
PS0 COMMITMENT P50 recognizes the importance for timely and effective isolation of the containment under accident conditions. P50 will review the design of BFS to assure that the final design provides for:
PS0 COMMITMENT P50 recognizes the importance for timely and effective isolation of the containment under accident conditions. P50 will review the design of BFS to assure that the final design provides for:
: 1. Diversity in the parameters sensed for the initiation of containment isolation, in accordance with SRP 6.2.4;
1.
: 2. Automatic isolation of non-essential systems upon containmer,t isolation signal;
Diversity in the parameters sensed for the initiation of containment isolation, in accordance with SRP 6.2.4; 2.
: 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.
Automatic isolation of non-essential systems upon containmer,t isolation signal; 3.
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 information on the definition of essential and non-essential systems.
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.
78
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 information on the definition of essential and non-essential systems. 78


NRR Lessons learned Task Force Short-Term Recommendations TITLE: Dedicated Penetrations for External Recombiners or Post-Accident Purce Systems Section 2.1.5.a     .
NRR Lessons learned Task Force Short-Term Recommendations TITLE: Dedicated Penetrations for External Recombiners or Post-Accident Purce 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.
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 containment of each unit. This position is therefore not applicable.
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.
NRR Lessons learned Task Force Short-Term Recommendations TITLE: hertina BWR Containments (Section 2.1.5.b).
NRR Lessons learned Task Force Short-Term Recommendations TITLE: hertina BWR Containments (Section 2.1.5.b).
NRC STAFF POSITION It shall be required that the Vemont 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.
NRC STAFF POSITION It shall be required that the Vemont Yankee and Hatch 2 Mark I BWR contain-ments be inerted in a manner similar to other operating BWR plants.
Black Fox Station is designed with a Mark III Containment. This position is not applicable.
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. /b,0
b,0


NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant (Section 2.1.5.c).
NRR Lessons Learned Task Force Short-Term Recommendations TITLE:
Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant (Section 2.1.5.c).
NRC STAFF POSITION (Minority View).
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 an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.
1.
: 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 T?il-2.
All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.
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.
2.
l
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 T?il-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. l


NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Integrity of Systems Outside Containment likely to Contain Radioactive Materials (Encineered Safety Systems and Auxiliary Systems   for PWR's and BWR's (Section 2.1.6.a).
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Integrity of Systems Outside Containment likely to Contain Radioactive Materials (Encineered 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 e program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as practical levels. This program shall include the following:
NRC STAFF POSITION Applicants and licensees shall immediately implement e program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as practical levels. This program shall include the following:
: 1. Imediate Leak Reduction.
1.
: a. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
Imediate Leak Reduction.
: b. Measure actual leakage rates with system in operation and report them to the NRC.
Implement all practical leak reduction measures for all systems a.
: 2. Continuing Leak Reduction.
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.
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.
PS0 COF"41TMENT 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 oreventive 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.
PS0 COF"41TMENT 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 oreventive 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 wil1 contain the results of the above desig" and ope.ations review.
The FSAR wil1 contain the results of the above desig" and ope.ations review.
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, g 2-


NRR Lessons Learned Task Force Shcrt-Term Recorrendations TITLE: Design Review of Plant Shielding of Spaces for Post-Accident Operations Section 2.1.6.b .
NRR Lessons Learned Task Force Shcrt-Term Recorrendations 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-accide'it 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 ue unduly limited or safety equipment may be 0.' duly degr.ided by the radiation fields during post-accident operations of these systems.
NRC STAFF POSITION With the assumption of a post-accide'it 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 ue unduly limited or safety equipment may be 0.' duly degr.ided by the 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.
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.
P50 COMMITMDE PS9 recognizes, as a result of the TMI-2 event, the need to assure necessary access to vital areas and protecticn 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 described in Regulatory Guide 1.3, will evaluate the BFS design for unacceptable limitations on personnel access and occupancy or undu" degradation of 2fety-related equipment curing post-acciocct operations. Thr evaluation will consider alternatives, including layout changes, increased use of permanent shielding, temporary shielding, or proce.iural controls.
P50 COMMITMDE PS9 recognizes, as a result of the TMI-2 event, the need to assure necessary access to vital areas and protecticn 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 described in Regulatory Guide 1.3, will evaluate the BFS design for unacceptable limitations on personnel access and occupancy or undu" degradation of 2fety-related equipment curing post-acciocct operations. Thr evaluation will consider alternatives, including layout changes, increased use of permanent shielding, temporary shielding, or proce.iural controls.
The The evaluation will determine changes needed throughout Black Fox Station.
The The evaluation will determine changes needed throughout Black Fox Station.
results of the evaluation and a description of the changes will be reflected in the final design presented in the FSAR.
results of the evaluation and a description of the changes will be reflected in the final design presented in the FSAR. o
o


NRR Lessons Learned Task Force Short-Term Recorrendations TITLE: Automatic Initiation of the Auxiliary Feedwater System for PWR's (Section 2.1.7.a) .
NRR Lessons Learned Task Force Short-Term Recorrendations TITLE: Automatic Initiation of the Auxiliary Feedwater System for PWR's (Section 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.
This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.
                                                '5-
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                                    ',  lb o


NRR Lessons Learned Task Force Short-Tern Recorrendations TITLE: Auxiliary Feedwater Flow Indication to Steam Generators for PWR's Section 2.1.7.b   .
NRR Lessons Learned Task Force Short-Tern Recorrendations TITLE: Auxiliary Feedwater Flow Indication to Steam Generators for PWR's Section 2.1.7.b This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.
This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.
NRR Lessons Learned Task Force Short-Term Recommendations TITLE:
NRR Lessons Learned Task Force Short-Term Recommendations TITLE:   Improved Post-Accident Samplino Capability (Section 2.1.8.a).
Improved Post-Accident Samplino 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 detemine the capability of personnel to promptly obtain (less than 1 hour) 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.
NRC STAFF POSITION A design and operational review of the reactor coolant and containment atmos-phere sampling systems shall be performed to detemine the capability of personnel to promptly obtain (less than 1 hour) 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.
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 spectrun analysis facilities shall be performed to determine the capability to promptly (less than 2 hcurs) 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 reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
A design and operational review of the radiological spectrun analysis facilities shall be performed to determine the capability to promptly (less than 2 hcurs) 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 reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be perforemd in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.
If the review indicates that the analyses required cannot be perforemd in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boren and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed prcmptly; i.e. the boron sample analysis within an hour and the chloride sample analysis within a shift.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boren and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed prcmptly; i.e. the boron sample analysis within an hour and the chloride sample analysis within a shift.
PS0 COM"ITMENT PS0 will perform a design and cperational review of the reactor coolant and con-tainment atmospheric sampling system, the radioisotope analysis facilities, and chemical analyses to achieve prompt and safe sample acquisition and analysis in accordance with the position stated above. Results of these studies will be presented in the FSAR.
PS0 COM"ITMENT PS0 will perform a design and cperational review of the reactor coolant and con-tainment atmospheric sampling system, the radioisotope analysis facilities, and chemical analyses to achieve prompt and safe sample acquisition and analysis in accordance with the position stated above. Results of these studies will be presented in the FSAR.
NRR Lessons Learned Task Force Short-Tem Recomendations TITLE: Increased Range of Radiation Monitors (Section 2.1.8.b).
NRR Lessons Learned Task Force Short-Tem Recomendations TITLE:
Increased Range of Radiation Monitors (Section 2.1.8.b).
NRC STAFF p0SITION The requirements associated with this recommendation should be considered as advanced implementation of certain requiret.2nts to be included in a revision to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident,"
NRC STAFF p0SITION The requirements associated with this recommendation should be considered as advanced implementation of certain requiret.2nts 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.
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 oeprating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.
1.
: a. Noble gas effluent monitors with an upper range capacity of 10 5 uCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal oeprating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.
: b. Noble gas effluent monitoring shall be provided for the' total range of concentration extending from a minimum of 10-7 uti/cc (Xe-133) to a maximum of 105 uCi/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors shall overlap by a factor of ten.
5 a.
: 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 provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.
Noble gas effluent monitors with an upper range capacity of 10 uCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
: 3. In-containment radiation level monitors with a maximum range of 10 8 rad /hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment.
b.
Noble gas effluent monitoring shall be provided for the' total range of concentration extending from a minimum of 10-7 uti/cc (Xe-133) to a maximum of 105 uCi/cc (Xe-133). Multiple monitors are considered to be necessary 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 provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.
8 3.
In-containment radiation level monitors with a maximum range of 10 rad /hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment.
pSC COMMITMENT PS0 shall provide the monitors as required in the staff position, and will dccument a description of the same in the FSAR.
pSC COMMITMENT PS0 shall provide the monitors as required in the staff position, and will dccument a description of the same in the FSAR.
7 NRR Lessons Learned Task Force Short-Term Reco= endations TITLE:   Improved In-Plant Iodine Instrumentation (Section 2.1.8.c).
7 NRR Lessons Learned Task Force Short-Term Reco= endations 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.
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 COMMITPENT PS0 will provide instrumentation, training of personnel and the technical procedures for accurately determining airborne iodine concentration throughout the plant ur.de" accident conditions, with documentation to be provided in the FSAR.
PS0 COMMITPENT PS0 will provide instrumentation, training of personnel and the technical procedures for accurately determining airborne iodine concentration throughout the plant ur.de" accident conditions, with documentation to be provided in the FSAR. 86
86


NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Analysis of Desien and Off-Normal Transients and Accidents Section 2.1.9).
NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Analysis of Desien and Off-Normal Transients and Accidents Section 2.1.9).
NRC STAFF POSITION Analyses, procedures, and training addressing the following are required:
NRC STAFF POSITION Analyses, procedures, and training addressing the following are required:
: 1. Samil break loss-of-coolant accidents;
1.
: 2. Inadequate core cooling; and
Samil break loss-of-coolant accidents; 2.
: 3. Transients and accidents.
Inadequate core cooling; and 3.
Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Tap 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 emergency proce-dures and in support of an elentual long-term verification of cogliance with Appendix K of 10 CFR Part 50.
Transients and accidents.
Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Tap 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 emergency proce-dures and in support of an elentual long-term verification of cogliance 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:
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:
1.
Low reactor coolant system inventory (two examples will be required:
LOCA with forced flow; LOCA without forced flow);
LOCA with forced flow; LOCA without forced flow);
: 2. Loss of natural circulation (due to loss of heat sink).
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 approached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real tire far enough that all important phenomena and instrument indications are included.
These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real tire far enough that all important phenomena and instrument indications are included.
Each case should then be repeated taking credit for correct operator action.
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 coerators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3b in this appendix).
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 coerators 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.
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 to 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 present, 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 ccmplete loss of auxiliary feedwater Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--
Consequential failures shall also be considered. Failures of the operators to 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 present, 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 ccmplete loss of auxiliary feedwater Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--
Continued.
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.
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.
Line 200: Line 263:
The transient and accident analyses shr.11 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 infomation not available from an event tree.
The transient and accident analyses shr.11 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 infomation 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 more serious accidents.
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 more serious accidents.
The information derived 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.
The information derived 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 trane!ents should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for ccmparisons 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 rathods being used to generate emergency procedures.
In addition to analyses performed by the reactor vendors, analyses of selected trane!ents should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for ccmparisons 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 rathods 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 operaticn of the Black Fox Station.
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 operaticn of the Black Fox Station.
PS0 understands that analysis and emergency procedures or guidelines for:
PS0 understands that analysis and emergency procedures or guidelines for:
: 1. Smdl break loss-of-coolant accidents;
1.
: 2. Inadequate core cooling; and C)0 Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--
Smdl break loss-of-coolant accidents; 2.
Inadequate core cooling; and C)0 Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)--
Continued.
Continued.
: 3. Transients and accidents are being generated by the operating Boiling Water Reactor Owners' Group in response to the Bulletins and 0 der Task Force. These analyses are being generalized first to cover BWR/1-5 type power plants and will be extended by General Electric Copany to cover the BWR/6 System generically. Each of the specific requirements stated in the above position tue been identified by the Bulletins and Orders Task Force. As this asscssnent 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 amergency procedures guidelines, and these guidelines will be implemented in the Black Fox Station procedures and training programs as appropriate.
3.
c,
Transients and accidents are being generated by the operating Boiling Water Reactor Owners' Group in response to the Bulletins and 0 der Task Force. These analyses are being generalized first to cover BWR/1-5 type power plants and will be extended by General Electric Copany to cover the BWR/6 System generically. Each of the specific requirements stated in the above position tue been identified by the Bulletins and Orders Task Force. As this asscssnent 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 amergency procedures guidelines, and these guidelines will be implemented in the Black Fox Station procedures and training programs as appropriate. c,


NRR Lessons Le:rned Task Force Short-Term Recommendations TITLE: Shift Supervisor's Responsibilities (Section 2.2.1.a).
NRR Lessons Le:rned Task Force Short-Term Recommendations TITLE: Shift Supervisor's Responsibilities (Section 2.2.1.a).
NRC STAFF POSITION
NRC STAFF POSITION 1.
: 1. The highest level of corporate management of each licensee shall 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.
The highest level of corporate management of each licensee shall 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 room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel . Particular emphasis shall be placed on the following:
2.
: 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 on duty in the control room. The idea shall be reinforced that 1he shift supervisor should not become totally involved in any single operation in times of emergency when mLltiple operations are required in the control room.
Plant procedures shall be reviewed to assure that the duties, responsi-bilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:
: b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct he activities of control rtom operators. Persons authorized to rt ieve the shift supervisor shall be specified,
a.
: c. If the shift supervisor is temporar.ily absent from the control room during routine operatior.3, a lead control room operator shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified.
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 on duty in the control room. The idea shall be reinforced that 1he shift supervisor should not become totally involved in any single operation in times of emergency when mLltiple operations are required in the control room.
: 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.
b.
: 4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each Jtility responsible for plant operations. Admini-strative functions that detract from or are subordinate to the management responsibility for assu ing the safe operation mf the plant shall be delegated to other operations personnel not on duty in the control room.
The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct he activities of control rtom operators.
c4
Persons authorized to rt ieve the shift supervisor shall be specified, c.
If the shift supervisor is temporar.ily absent from the control room during routine operatior.3, a lead control room 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 Jtility responsible for plant operations. Admini-strative functions that detract from or are subordinate to the management responsibility for assu ing the safe operation mf the plant shall be delegated to other operations personnel not on duty in the control room. c4


Shif t Supervisor's P.esponsibilities (Section 2.2.la)--
Shif t Supervisor's P.esponsibilities (Section 2.2.la)--
Continued.
Continued.
pSO COMMITMENT PS0 comits 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 plant 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.
pSO COMMITMENT PS0 comits 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 plant 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. g3
g3


NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Shift Technical Advisor (Section 2.2.1.b).
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Shift Technical Advisor (Section 2.2.1.b).
Line 234: Line 304:
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Shift and Relief Turnover Procedures (Section 2.2.1.c).
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:
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 complete and sign.
1.
A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign.
The following items, as a minimum, shall be included in the checklist:
The following items, as a minimum, shall be included in the checklist:
: a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist);
Assurance that critical plant parameters are within allowable a.
: b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients anr' =idents by a check of the control console (what to check ano criteria for acceptable status shall be included on the checklist);
limits (parameters and allowable limits shall be listed on the checklist);
: c. Identification of systems and components that are in a degraded mode of operation permitted 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 separa+e entry on the checklist).
b.
: 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
Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients anr' =idents by a check of the control console (what to check criteria for acceptable status shall be included on the checklist);
: 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).
ano Identification of systems and components that are in a degraded c.
PSO COMMITENT PS0 commits to compliance with the above position and concurs that it is a prudent management approach to plant operations.
mode of operation permitted by the Technical Specifications.
95
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 separa+e 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).
PSO COMMITENT PS0 commits to compliance with the above position and concurs that it is a prudent management approach to plant operations. 95


NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Control Room Access (Section 2.2.2.a).
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 roem operators),
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 roem operators),
to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel. Provisions shall include the following:
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 administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access;
1.
: 2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event af an emergency. The line of succession for the person in charge of tre control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall claarly 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.
Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access; 2.
PS0 COMMITMENT PSO will comply fully with this position and recognizes the importance of access control to the control room.
Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event af an emergency. The line of succession for the person in charge of tre control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall claarly 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.
N/
PS0 COMMITMENT PSO will comply fully with this position and recognizes the importance of access control to the control room. N/)D
                                      )D


NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Onsita Technical Support Center (Section 2.2.2.b).
NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Onsita Technical Support Center (Section 2.2.2.b).
Line 257: Line 331:
The licensee shall revise his emergency plans as necessary to incorporate the role and location of the tichnical support center.
The licensee shall revise his emergency plans as necessary to incorporate the role and location of the tichnical 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 6ccessible to the technical support center under emergency conditions. These documents shall incluce, 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).
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 6ccessible to the technical support center under emergency conditions. These documents shall incluce, 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 COMMITMENT An onsite technical support center as described above will bc             -sith the capability to display necessary plant status information for 1,i       is who are knoveledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to 'he 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.
PS0 COMMITMENT An onsite technical support center as described above will bc
A description of this center will be provided in the FSAR.
-sith the capability to display necessary plant status information for 1,i is who are knoveledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to 'he 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.
9'7
A description of this center will be provided in the FSAR. 9'7


NRR Lessens Learned _ Task Force Short-Term Recommendations TITLE: Onsite Operational Support Center (Section 2.2.2.c).
NRR Lessens Learned _ Task Force Short-Term Recommendations TITLE: Onsite Operational Support Center (Section 2.2.2.c).
NRC STAFF POSIT 7p fn 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 o which the operations support personnel will report in an emergency situation.
NRC STAFF POSIT 7p fn 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 o which the operations support personnel will report in an emergency situation.
omunications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of comunication and ranagement.
omunications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of comunication and ranagement.
PS0 C0KMTMENT PS0 will designate an area to serve as the operational support center as described in the above position. The support canter will be physically separated from the control room, and appropriate comun1 cation 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 Suppcrt Center, as well as establish the methods and lines of communication and management control . The location of the Center will be provided in the FSAR.
PS0 C0KMTMENT PS0 will designate an area to serve as the operational support center as described in the above position. The support canter will be physically separated from the control room, and appropriate comun1 cation 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 Suppcrt Center, as well as establish the methods and lines of communication and management control. The location of the Center will be provided in the FSAR. gg
gg


NRR Lessons Learned Task Force Short-Term' Recommendations TIi ': Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability (Section 2.2.3           .
NRR Lessons Learned Task Force Short-Term' Recommendations TIi ': 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 threshold of complete loss of safety function.
NRC STAFF p0SITION All NRC nuclear power plant licensees shall provide information to define a limiting operational condition based on a threshold of complete loss of safety function. Identification of a huran or operational error that prevents or could prevent the ai complishment of a safety function required by NRC regulations and analy7ej in the license application shall require placement of the plant in a hot sht.tdown ccidition within 8 hours and in a cold shutdown condition within 24 baurs.
Identification of a huran or operational error that prevents or could prevent the ai complishment of a safety function required by NRC regulations and analy7ej in the license application shall require placement of the plant in a hot sht.tdown ccidition within 8 hours and in a cold shutdown condition within 24 baurs.
The loss of operability of a saft-;v function shall include consideration of the ncessary instrumentation, cetrols, 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 loss of operability of a saft-;v function shall include consideration of the ncessary instrumentation, cetrols, 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:
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 hours;
1.
: 2. Within 24 hours, bring the plant to cold shutdown;
If the plant is critical, restore the safety function (if possible) and place the plant in a hot shutdown condition within 8 hours; 2.
: 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;
Within 24 hours, bring the plant to cold shutdown; 3.
: 4. Determine corrective actions and measures to prevent recurrence of the specific loss of operability for the particular safety function and generally for any safety function;
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.
: 5. Report the event within 24 hours by telephone and confirm by tele-graph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designee;
Determine corrective actions and measures to prevent recurrence of the specific loss of operability for the particular safety function and generally for any safety function; 5.
: 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.
Report the event within 24 hours by telephone and confirm by tele-graph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designee; 6.
                                                )h Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability (Section 2.2.3)--Continued.
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.
: 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.
)h Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability (Section 2.2.3)--Continued.
PS0 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 for under the Administrative Procedures Act. This approach was emphasized in Dr. Mattson's letter of July 18,1979 to Mr. Denton, attav     .. During the July 20 meeting with PS0, Mr. Denton stated that any commitment to the position must await the rulemaking process.
7.
In view of the foregoing, no commitment to the above position is required of .50 at this tire. PS0 does agree to comply with any requirement ultimately determined by the rulemaking.                                                                 .
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.
[60
PS0 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 for under the Administrative Procedures Act. This approach was emphasized in Dr. Mattson's letter of July 18,1979 to Mr. Denton, attav During the July 20 meeting with PS0, 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.50 at this tire. PS0 does agree to comply with any requirement ultimately determined by the rulemaking.
. [60


ga ascg                                 UMTED STATES 6,.         fg y   3 , ,(g g               NUCLEAR REGUL ATORY COMMISSION 3g    7        ;                      W ASHINGTON D. C. 20555 QQ.f July 18,1979 MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:               Roger J. Mattson, Director THI-2 Lessons Learned Task Force
ga ascg UMTED STATES 6,.
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QQ.f July 18,1979 MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:
Roger J. Mattson, Director THI-2 Lessons Learned Task Force TMI-2 LESSONS LEAR"ED TASK FORCE


==SUBJECT:==
==SUBJECT:==
TMI-2 LESSONS LEAR"ED TASK FORCE REPORT (SHORT TERM) NUREG-0578 Enclosed is the first report of the 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 appliutions. There are 23 specific recomenc'ations in 12 broad areas (nine in the area The 23ofrecom-design and analysis and three in the area of operations).
It contains a set of short term recommendations to be implemented in two stages over the next 18 months on operating plants, plants under There are construction, and pending construction permit appliutions.
mendations would provide substantial, additional protection which is required for the public health and safety.
23 specific recomenc'ations in 12 broad areas (nine in the area of design and analysis and three in the area of operations).
All but one of the 23 reco=endations have a majority concurrence by the Task Force. The excepi. ion 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.
The 23 recom-mendations would provide substantial, additional protection which is required for the public health and safety.
All but one of the 23 reco=endations have a majority concurrence by the The excepi. ion is the recommended requirement to provide Task Force.
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 recomends that this matter deserves further evaluation in conjunction with other hydrogen generation and control questions being reviewed by the Task Force for its final report.
majority of the Task Force recomends that this matter deserves further evaluation in conjunction with other hydrogen generation and control questions being reviewed by the Task Force for its final report.
Three of the recomendations appear to require changes in existing regulations for which the Task Force recommends immediately effective rulenaking. They are: 1) inerting of MKI 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 er operational error.
Three of the recomendations appear to require changes in existing regulations for which the Task Force recommends immediately effective
: 1) inerting of MKI and MK II BWR containments that rulenaking. They are:
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 er operational error.
to develop the required Commission papers and carry through with these rulemaking actions.
to develop the required Commission papers and carry through with these rulemaking actions.
The 23 recomended actions were discussed with the Regulatory Requirements Review Committee (June 22, 1979), the Commission (June 25,1979), the TMI-2 Subcommittee of the ACRS (July 11,1979), and the ACRS (July 12, 1979).
The 23 recomended actions were discussed with the Regulatory Requirements Review Committee (June 22, 1979), the Commission (June 25,1979), the 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.
TMI-2 Subcommittee of the ACRS (July 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.
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2 Harold R. Denton The Task Force recommends that time not be taken to request and evaluate public comments on these short tem requirements prior to their promulgation as licensing requirements or rules because they are safety significant matters that require prompt application to operating Otherreactors and Tl1I-2   operating accident     license review   groups applications in the late stages of review.
2 Harold R. Denton The Task Force recommends that time not be taken to request and evaluate public comments on these short tem requirements prior to their promulgation as licensing requirements or rules because they are safety significant matters that require prompt application to operating reactors and operating license Other Tl1I-2 accident review groups applications in the late stages of review.
and the Lessons Learned Task Force are continuing to evaluate the longer term implications of the accident. Any public coments on the short tem 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.
and the Lessons Learned Task Force are continuing to evaluate the longer term implications of the accident. Any public coments on the short tem 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 tem 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.
Having identified the 23 specific recomendations for short tem 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 span of the These longer Task Force) before other regulatory actions are recommended.
term interests of the Task Force are described in Section Three of the report. The Task Force intends to develop its final recommendations and issue a final report in early September 1979. The topics to be addressed 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 under construction. The Task Force does not believe that allowing new plants 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.
term interests of the Task Force are described in Section Three of the The Task Force intends to develop its final recommendations and report.
issue a final report in early September 1979. The topics to be addressed 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 11, 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 tem reco=endations 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.
On July 11, 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 tem reco=endations 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 by the various NRR organizations in the course of their review. Where appropriate, we made clarifying changes in the language of the report. The principal substantive change occurred in the fom and schedules of the implementation section (Appendix B). Some of the com ents addressed matters that the Task Force has deferred for consideration in its final report. There are significant differences of opinion within the staff on two of the Task Force reco=endations, as follows: a) the need for recommendation 2.2.3 concerning rulemaking for revised limiting conditions for operation (some agree with the recomendation and others prefer more stringent enforcement actions using existing regulatory machinery) and b) the need for the minority Task Force recommendation 2.1.5.c concerning rulemaking for backfit of recombiner capability (some support the minority recommendation, others do not).
We have reviewed and considered the detailed comments supplied by the various NRR organizations in the course of their review. Where appropriate, we made clarifying changes in the language of the report. The principal substantive change occurred in the fom and schedules of the implementation section (Appendix B). Some of the com ents addressed matters that the Task There are Force has deferred for consideration in its final report.
Having considered these co=ents and made changes to the report where appropriate to reconcile them with the intent of the Task Force, I reccmend that you:
significant differences of opinion within the staff on two of the Task Force a) the need for recommendation 2.2.3 concerning reco=endations, as follows:
: a. direct the i=ediate implementation by DPM, DOR or B&OTF, as appropriate, of all the short tem recommendations, except the three rulemaking matters, through the issuance of licensing positions to operating plant licensees, plants under construction, and construction permit applicants.
rulemaking for revised limiting conditions for operation (some agree with the recomendation and others prefer more stringent enforcement actions using existing regulatory machinery) and b) the need for the minority Task Force recommendation 2.1.5.c concerning rulemaking for backfit of recombiner capability (some support the minority recommendation, others do Having considered these co=ents and made changes to the report where not).
appropriate to reconcile them with the intent of the Task Force, I reccmend that you:
direct the i=ediate implementation by DPM, DOR or B&OTF, as appropriate, of all the short tem recommendations, except the three rulemaking a.
matters, through the issuance of licensing positions to operating plant licensees, plants under construction, and construction permit applicants.
( 0 2-
( 0 2-


3 Harold R. Denton
3 Harold R. Denton request the fomulation of immediately effective rules by the Office of Standards Development for action by the Comission on the three b.
: b. request the fomulation of immediately effective rules by the Office of Standards Development for action by the Comission on the three rulemaking matters.
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.
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 RogerJ.dattson,' Director TMI-2 Lessons Learned Task Force
p RogerJ.dattson,' Director TMI-2 Lessons Learned Task Force


==Enclosure:==
==Enclosure:==
as stated cc: Chaiman Hendrie
as stated cc: Chaiman Hendrie Comissioner Gilinsky Commissioner Kennedy Comissioner Bradford Comissioner Ahearne ACRS (20)
* Comissioner Gilinsky Commissioner Kennedy Comissioner Bradford Comissioner Ahearne ACRS (20)
Policy Evaluation SECY L. V. Gossick, EDO S. Levine, RES R. Minogue, SD V. Stello, IE M. Rogovin, Special Inquiry J. Fouchard, PA (20)
Policy Evaluation SECY L. V. Gossick, EDO S. Levine, RES R. Minogue, SD V. Stello, IE M. Rogovin, Special Inquiry J. Fouchard, PA (20)
C. Kamerer, CA (20) tiRC PDR
C. Kamerer, CA (20) tiRC PDR
                                    'O3
'O3


RESPONSE TO INSPECTION & ENFORCEMENT BULLETIN 79-08.
RESPONSE TO INSPECTION & ENFORCEMENT BULLETIN 79-08.
Line 323: Line 411:


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.
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 the apparent actions taken during the early phases of the accident; (2)(3) the necessity operational errors which led to eventual core damage; and to systematically analyn plant conditions and parameters and take appropriate corrective action;
This review should be directed toward understanding:
: b. Operational personnel should be instructed to: (1) not override autoratic 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 indicaticn when one or more confirratory indications are available;
(1) the extreme a.
: c. All licensed operators and Plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be documntedin plant records.
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)(3) the necessity the apparent operational errors which led to eventual core damage; and to systematically analyn plant conditions and parameters and take appropriate corrective action; Operational personnel should be instructed to:
(1) not override autoratic b.
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 indicaticn when one or more confirratory indications are available; All licensed operators and Plant management and supervisors with operational c.
responsibilities shall participate in this review and such participation shall be documntedin plant records.
PS0 C0??iITMENT public Service Company of Oklahoma has established a Technical Advisory Comittee (TAC) to assess the events at Three Mile Island, Unit 2, and to apply the lessons learned to its Black Fox Station Project. This comittee was established at the direction of the President and Chief Executive Officer of the Company and reports its findings and recomendations directly to the Review and Audit Ccmittee.
PS0 C0??iITMENT public Service Company of Oklahoma has established a Technical Advisory Comittee (TAC) to assess the events at Three Mile Island, Unit 2, and to apply the lessons learned to its Black Fox Station Project. This comittee was established at the direction of the President and Chief Executive Officer of the Company and reports its findings and recomendations directly to the Review and Audit Ccmittee.
These findings and recc=endations will then be implemented by the Review and audit Comittee.
These findings and recc=endations 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:
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 "Lesscns Learned Task Force"--NUREG-0578
USNRC's "Lesscns Learned Task Force"--NUREG-0578 The President's Cc= mission on Three Mile Island EPRI--Nuclear Sa fety Analysis Center
                . The President's Cc= mission on Three Mile Island EPRI--Nuclear Sa fety Analysis Center
[f IEB 79-08 Task 1--Continued.
[f
Generic vendor programs Atomic Industrial Forum TMI Policy Committee NRC Special Investigation (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.
IEB 79-08 Task 1--Continued.
It is aware of the activities of various other legislative and regulatory investigations and will assess future recomencations from them.
          . Generic vendor programs Atomic Industrial Forum TMI Policy Committee
The assessment 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.
          . NRC Special Investigation (Rogovin)
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 P50 continously informed of any new developments arising from the ongoing investigations by EPRI and other organizations.
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 recomencations from them.
The assessment 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 P50 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 precedures, staffing and training progrem, and emergency response plan incorporates the benefits of the TMI investigation to the fullest extent practicable.
The objective of the TAC and its consultants is to ensure that the Black Fox Sation design, construction, operating precedures, staffing and training progrem, 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.
The effort is directed toward understanding:
Prior to completion of operating procedures and training instructions for 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--Continvod.
(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.
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 available.   (See also commitments made under ICB 79-08 Task 5).
Prior to completion of operating procedures and training instructions for operation of the Black Fox Station, these procedures and instructions will be reviewed to assure that operational personnel are instructed to:
The Manager, Black Fox Station   4 the Manager, Nuclear Training are assigned to the TAC to ensure that operationa. 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 tc review administrative mechanisms to ensure that lessons learned are incorporated into the station training programs.
(1) not override IEB 79-08 Task 1--Continvod.
Findings and recorr.endations from the TAC will be documented in the Project files and conformance with each specified commitment will be incorporated into th c documentation system.
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 available.
(o9
(See also commitments made under ICB 79-08 Task 5).
The Manager, Black Fox Station 4 the Manager, Nuclear Training are assigned to the TAC to ensure that operationa. 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 tc review administrative mechanisms to ensure that lessons learned are incorporated into the station training programs.
Findings and recorr.endations from the TAC will be documented in the Project files and conformance with each specified commitment will be incorporated into th c documentation system. (o9


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 ranual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
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 ranual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
PSO 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.
PSO 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 cr ranual, and any necessary ranual actions will be covered by apr;opriate procedures.
This isolation may be automatic cr ranual, and any necessary ranual actions will be covered by apr;opriate procedures. l06
l06


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 tranual action necessary, describe in sunnary form the procedure by which this action is, taken in a timely sense.
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 tranual action necessary, describe in sunnary form the procedure by which this action is, taken in a timely sense.
PS0 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 systers that are used when the rain feedwater system is not operable will be reviewed. Both automatic and ranual actions will be assessed for adequacy, and any necessary manual actions will be addressed by procedures to assure timely actuations.
PS0 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 systers that are used when the rain feedwater system is not operable will be reviewed. Both automatic and ranual actions will be assessed for adequacy, and any necessary manual actions will be addressed by procedures to assure timely actuations.
(09
. (09


IEB 79-08 Task 4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems. Describe other redundant instrumentation which the operator might have to give the same infomation regarding plant status. Instruct operators to utilize other available information to initiate safety systems.
IEB 79-08 Task 4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems. Describe other redundant instrumentation which the operator might have to give the same infomation regarding plant status.
p50 COM'11TMENT 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 automatic and manual initiation of safety systems will be reviewed. Redundant instrumentation which the operator will have to give the same vessel level indications will be identified and factored into operator training, instruction, and procedures.
Instruct operators to utilize other available information to initiate safety systems.
lIO
p50 COM'11TMENT 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 automatic and manual initiation of safety systems will be reviewed.
Redundant instrumentation which the operator will have to give the same vessel level indications will be identified and factored into operator training, instruction, and procedures. lIO


IEB 79-08 Task 5 Revit i the action directed by the operating procedures and training instructions to ensure that:
IEB 79-08 Task 5 Revit i 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);
Operators do not override automatic actions of engineered safety a.
: b. Operators are provided additional information and instructions to not rely upon vcssel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.
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 vcssel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.
P50 COMMITMENT Prior to completion of operatin; procedures and training instructions, actions directed by tt.:se ir.structions will be reviewed to ensure that:
P50 COMMITMENT Prior to completion of operatin; procedures and training instructions, actions directed by tt.:se ir.structions will be reviewed to ensure that:
: a. Operators are directed not to override automatic action of engineered safety features enless continued operation of engineered safety features will result in unsafe plant conditions;     .
Operators are directed not to override automatic action of engineered a.
: b. Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual acticns, but to also examine other plant parameter indications in evaluating plant conditiens.
safety features enless continued operation of engineered safety features will result in unsafe plant conditions; b.
fll
Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual acticns, but to also examine other plant parameter indications in evaluating plant conditiens. fll


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 featur?s. 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 follcwing necessary ranipulations and are raintained in their proper positions dt ring all operational modes.
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 featur?s. 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 follcwing necessary ranipulations and are raintained in their proper positions dt ring all operational modes.
pSO COMMITMENT At the time of final design, i.e., FSAR submittal, PS0 will review ali safety-related valve positioning requirements and positive cont ols to assure that valves remain positioned in a manner to ensure the proper operation of engineered safety features.
pSO COMMITMENT At the time of final design, i.e., FSAR submittal, PS0 will review ali safety-related valve positioning requirements and positive cont ols 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 procedures for raintenance, 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 ranipulations and are maintained in the proper position during all operational modes.
In addition, prior to completion of related procedures, the procedures for raintenance, 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 ranipulations and are maintained in the proper position during all operational modes. ( (1
( (1


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.
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 sur.h an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate:
In particular, ensure that sur.h 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;
a.
: b. Whether such systems are isolated by the containment isolation signal;
Whether interlocks exist to prevent transfer when high radiation indication exists, and; b.
: c. The basis on which continued operability of the above features is assured.
Whether such systems are isolated by the containment isolation signal; c.
The basis on which continued operability of the above features is assured.
pS0 COM'4ITMENT 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 other release of radioactive gases and liquids will not occur inadvertently.
pS0 COM'4ITMENT 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 other 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.
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:
Each of the above systems will be reviewed to assure that:
: a. Interlocks exist to prevent transfer when high radiation indication exists, and;
a.
: b. Such systems are isolated by the containment isolation signal.
Interlocks exist to prevent transfer when high radiation indication exists, and; b.
                                              -4 0 -
Such systems are isolated by the containment isolation signal.
-4 0 -
ff3
ff3


IEB 79-08 Task 8 Review and modify as necessary your maintenance and test procedures to er.sure that they require:
IEB 79-08 Task 8 Review and modify as necessary your maintenance and test procedures to er.sure 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;
a.
: b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing:
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.
: c. Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.
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.
PSO COMMITMENT Prior to their completion, maintenance and test procedures for safety-related systems will be reviewed to ensure that they require:
PSO 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;
a.
: b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing;
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.
: c. Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from or returned to service.
Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing; c.
ik
Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from or returned to service. ik


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 connunication channel shall be established and maintained with NRC.
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 connunication channel shall be established and maintained with NRC.
PS0 COMMITMENT Prior to completion of the emergency plan and implementing procedures, NRC notification shall be incorporated to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
PS0 COMMITMENT Prior to completion of the emergency plan and implementing procedures, NRC notification shall be incorporated 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 the time of NRC notification, an open continuous connunication channel will be established and maintained with NRC.
Further, at the time of NRC notification, an open continuous connunication channel will be established and maintained with NRC. fl5
fl5


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.
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.
Line 405: Line 501:


IEB 79-08 Task 11 Propose changes, as required, to those technical specifications which must be modified as a result of your implementing the items above.
IEB 79-08 Task 11 Propose changes, as required, to those technical specifications 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.
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. II7
II7


RESPONSE TO SELECTED ISSUES ON EMERGENCY PREPAREDNESS t r
RESPONSE TO SELECTED ISSUES ON EMERGENCY PREPAREDNESS t
r


Emergency Preparedness
Emergency Preparedness 1.
: 1. Reculatory Guide 1.101 Emergency Planning For Nuclear Power Plants.
Reculatory Guide 1.101 Emergency Planning 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.
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 regulatory staff have indicated i. hat revisions to the uniform action level criteria will be forthcoming as a result of the experiences at TMI. PS0 will utilize these criteria in development of the BFS Emergency Plan.
Discussions with the regulatory staff have indicated i. hat revisions to the uniform action level criteria will be forthcoming as a result of the experiences at TMI. PS0 will utilize these criteria in development of the BFS Emergency Plan.
ii. Improved Samoling and Instrumentation Cacability.
ii.
These issues are covered in NUREG-0578 TMi-2 Lessons Learned Task Force Status Recort and Short-Term Recommendations as issue 2.1.8.     PS0 has addressed these requirements in our response to that section.
Improved Samoling and Instrumentation Cacability.
These issues are covered in NUREG-0578 TMi-2 Lessons Learned Task Force Status Recort and Short-Term Recommendations as issue 2.1.8.
PS0 has addressed these requirements in our response to that section.
iii. Emercency Operatina Center.
iii. Emercency Operatina 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. Appros       'e plant status and meteorological 6ta will be read directly from instrumentation ) laced in the 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,
iv. Imoroved Offsite Monitoring Capability.
'e plant status and meteorological and federal authorities involved. Appros 6ta will be read directly from instrumentation ) laced in the center.
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 Emerger;cy Perparedness - iv.   (Continued).
iv.
possible use of continuous radiation monitors with remote readout. PS0 also ccmits to closely monitor forthcoming regulatory guidance in this area to assure that appropriate capabilities are promptly factored into the BFS design and operation plan.
Imoroved Offsite Monitoring Capability.
: v. Adequacy of Protecti te Action Planning.
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 Emerger;cy Perparedness - iv.
(Continued).
possible use of continuous radiation monitors with remote readout.
PS0 also ccmits 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 Protecti te 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 Dapartment 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.
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 Dapartment 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 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 Comission, explains the State's status in preparing such a plan, and receiving NRC approval.
The attached {{letter dated|date=June 20, 1979|text=letter dated June 20, 1979}} from George Nigh, Governor, State of Oklahoma, to Joseph Hendrie, Chairman, U. S. Nuclear Regulatory Comission, 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 and submittal to 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 and submittal to NRC approval.
Concurrently, PS0 is establishing target tasks for the BFS Emergency Response Plan development. The plan will be submitted with the CSAR in support of the application for operating licenses.
Concurrently, PS0 is establishing target tasks for the BFS Emergency Response Plan development. The plan will be submitted with the CSAR in support of the application for operating licenses.
Our understanding from recent discussions 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 4,00u meters as reflected in the BFS Preliminary Safety Analysis Report and Environmental Report.
Our understanding from recent discussions 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 4,00u meters as reflected in the BFS Preliminary Safety Analysis Report and Environmental Report.
3ccordingly, we have reviewed the applicable discussion from the ER (@ 2.1.3.1) on the popluation projections within a ten-mile radius of the site. Also studied were PSAR tabluations of regional incorporated community statistics and population
3ccordingly, we have reviewed the applicable discussion from the ER (@ 2.1.3.1) on the popluation projections within a ten-mile radius of the site. Also studied were PSAR tabluations of regional incorporated community statistics and population
                                            -4E-
-4E-


Emergency Preparedness - v.   (Continued).
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.
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.
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 tiew Tulsa (eight miles WSW),
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 tiew Tulsa (eight miles WSW),
Fair Oaks (nine miles WiiW), 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.
Fair Oaks (nine miles WiiW), 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 ten-mile radius of the Black Fox Station is small--less than 15,000 in 1980 and less than 24,000 in 2020.
The accompanying ER Table 2-1-1 shows that the overall population density within the ten-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 In addition, such as state highways 18 and 33 and U. S. Highway 69 are identified.
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.
since Oklahoma is uniformly divided into square mile sections, each of the perpendicular lines forming uniform squares on the figure represents a transportation route.
In addition, since Oklahoma is uniformly divided into square mile sections, each of the perpendicular lines forming uniform squares on the figure represents a transportation route.
As a result of our review, we have ccncluded that implementation of protective measuras such'as evacuation is feasible over the lifetime of the station based on population estimates and evacuation routes.
As a result of our review, we have ccncluded that implementation of protective measuras such'as evacuation is feasible over the lifetime of the station based on population estimates and evacuation routes.
vi. Periadic Testina.
vi.
PS0 comments to periodically corduct local emergency plan testing to assure that
Periadic Testina.
[7d
that PS0 comments to periodically corduct local emergency plan testing to assure [7d


Emergency Preparedness - iv.   (Continued).
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.
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.
e
e
                                              -4 3-ILZ
-4 3-ILZ


    ,p.                                        STATE OF OKL AHOM A
STATE OF OKL AHOM A
: m. . . ,
,p.
7 1'".0 *1 OFFIG OF THE GOVERNOR g/ J                             212 STATE CAPITOL BUILDING s
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7 ".0 *1 OFFIG OF THE GOVERNOR g/ J 212 STATE CAPITOL BUILDING 1'
  -(" ..l,',,                               OKLAHOMA CITY. OKLAHOMA 73105 GEORGE NIGH June 2,0, 1979 45 / $21234.5 Mr. Joseph M. Hendrie, Chairman U. S. fluclear Regulatory Comission Washirgton, D. C. 20555
-("..
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l,',,
OKLAHOMA CITY. OKLAHOMA 73105 GEORGE NIGH June 2,0, 1979 45 / $21234.5 Mr. Joseph M. Hendrie, Chairman U. S. fluclear Regulatory Comission Washirgton, D. C.
20555


==Dear Mr. Hendrie:==
==Dear Mr. Hendrie:==
 
I share your concern with regard to states having ade-quate radiological emergency response plans in opera-tion which support fixed nuclear facilities.
I share your concern with regard to states having ade-quate radiological emergency response plans in opera-tion which support fixed nuclear facilities. I appreciate your kind offer to assist in preparing such a plan through the mechanism of the Federal Interagency Re-gional Advisory Connittee and your agency.
I appreciate your kind offer to assist in preparing such a plan through the mechanism of the Federal Interagency Re-gional Advisory Connittee and your agency.
-                      The Occupational r nd Radiological Health Service of the Oklahoma Depa.tment of Health, in cooperation with the Oklahoma Office of Civil Defense, has recently completed a preliminary draft of Oklahcma's radiologi-cal plan. Copies of this draft have been circulated to my office, several State executive agencies, the tiRC Office of State Programs, and Public Service Company of Oklahoma for comments. Following revision in accord with these comments, the plan will be circulated for       -
The Occupational r nd Radiological Health Service of the Oklahoma Depa.tment of Health, in cooperation with the Oklahoma Office of Civil Defense, has recently completed a preliminary draft of Oklahcma's radiologi-cal plan. Copies of this draft have been circulated to my office, several State executive agencies, the tiRC Office of State Programs, and Public Service Company of Oklahoma for comments. Following revision in accord with these comments, the plan will be circulated for comment to these State agencies,' local officials, the public, and the tiRC. Our current schedule calls for a final version of the plan to be ready by early 1980.
comment to these State agencies,' local officials, the public, and the tiRC. Our current schedule calls for a final version of the plan to be ready by early 1980. He fully intend and expect to receive iiRC concurrence to the final plan several years prior to the now anticipated operational status of the Black Fox Station in 1985.
He fully intend and expect to receive iiRC concurrence to the final plan several years prior to the now anticipated operational status of the Black Fox Station in 1985.
interel r yours, c
interel r yours, c
                      ,          U George ligh t       e,
U George ligh t
e,


BFS 2.1.3.1 Population Within 10 Miles. A map of the 10-mile area of the BFS
BFS 2.1.3.1 Population Within 10 Miles. A map of the 10-mile area of the BFS
(
(
Site is presented on Figure 2.1-6. The map is overlayed with concentric I
Site is presented on Figure 2.1-6.
circles, centered on the central plant complex with radii of 1, 2, 3, 4, 5, and 10 =11es, and with radial lines forming 22-1/2 degree sectors centered on the 16 cardinal conpass points. Table 2.1-1 presents the corresponding pro-jected residential population within each annular and radial sector segnents for the expected first year of plant operation (1983) and by census decade beginning with 1990 through the end of the anticipated plant life (2020).
The map is overlayed with concentric circles, centered on the central plant complex with radii of 1, 2, 3, 4, 5, I
and 10 =11es, and with radial lines forming 22-1/2 degree sectors centered on the 16 cardinal conpass points. Table 2.1-1 presents the corresponding pro-jected residential population within each annular and radial sector segnents for the expected first year of plant operation (1983) and by census decade beginning with 1990 through the end of the anticipated plant life (2020).
The largest cu=ulative population density for this area through the year 2020, occurs within the 4-=ile 1.dius area, in which the town of luola is located.
The largest cu=ulative population density for this area through the year 2020, occurs within the 4-=ile 1.dius area, in which the town of luola is located.
The 10-mile radius area is pri=arily rural and is expected to re=ain as such during the period of plant operation. Base data and methodology of population projections are presented in Subsection 6.1.4.2.
The 10-mile radius area is pri=arily rural and is expected to re=ain as such during the period of plant operation. Base data and methodology of population projections are presented in Subsection 6.1.4.2.
Line 473: Line 585:
* 120376 t
* 120376 t
7
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                                              ')
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BFS The town of inola is the only significant population concentration with-(
BFS The town of inola is the only significant population concentration with-(
in the 10-mile area. The IQ74 esti=ated population of Inola is 1176 with projections presented in the Co== unity Development Plan, Inola Oklahe=a, in-creasing to 4200 by the year 2000 (6). There are three other small co== unities wA:hin 10 miles of BFS in addition to the town of Inola. The other co== unities are New Tulsa (8 miles WSW), Fair Oaks (9 miles WNW), and Tiawah (10 miles N).
The IQ74 esti=ated population of Inola is 1176 with in the 10-mile area.
projections presented in the Co== unity Development Plan, Inola Oklahe=a, in-creasing to 4200 by the year 2000 (6). There are three other small co== unities wA:hin 10 miles of BFS in addition to the town of Inola. The other co== unities are New Tulsa (8 miles WSW), Fair Oaks (9 miles WNW), and Tiawah (10 miles N).
New Tulsa and Fair Oaks are incorporated entities in Wagoner County while Tiawah is unincorporated and located in Rogers County. New Tulsa and Fair Oaks populations are not expected to increase significantly according to projections by the Ohiaho=a Employ =ent Security Co==ission (7). Much of the Tiawah current, estimated population of 95, is located beyond the 10-mile radius (8).
New Tulsa and Fair Oaks are incorporated entities in Wagoner County while Tiawah is unincorporated and located in Rogers County. New Tulsa and Fair Oaks populations are not expected to increase significantly according to projections by the Ohiaho=a Employ =ent Security Co==ission (7). Much of the Tiawah current, estimated population of 95, is located beyond the 10-mile radius (8).
2.1.3.2   Pooulation Jetween 10 and 50 Miles. Figure 2.1-7 shows the region within 50 miles of the reactor locations in northeast Oklahoma with concentric circles drawn at 10-mile radius intervals and wi:S radial lines defining sectors centered on the 16 cardinal compass firections. The projected pcpu-lations for 1983,1990, 2000, 2010, and 202C for each annular and radial sector segments are presented in Table 2.1-2. The =ethods for estimating population distribution are descri'aed in subsection 6.1.4.2. The nearest population center (as defined in 10 CFR 100) at the tire of startup of Unit 1 is Tulsa, Oklahoma with a 1970 census population of 330,350 (9). The nearer boundary of the densely populated area of Tulsa deter =ined by inter-pretat;     of July 1974 serial pnotographs is located 13 miles west of the Site. This distance is 5.2 times the low population zone radius of 2.5 miles.
2.1.3.2 Pooulation Jetween 10 and 50 Miles.
Figure 2.1-7 shows the region within 50 miles of the reactor locations in northeast Oklahoma with concentric circles drawn at 10-mile radius intervals and wi:S radial lines defining sectors centered on the 16 cardinal compass firections. The projected pcpu-lations for 1983,1990, 2000, 2010, and 202C for each annular and radial sector segments are presented in Table 2.1-2.
The =ethods for estimating population distribution are descri'aed in subsection 6.1.4.2.
The nearest population center (as defined in 10 CFR 100) at the tire of startup of Unit 1 is Tulsa, Oklahoma with a 1970 census population of 330,350 (9). The nearer boundary of the densely populated area of Tulsa deter =ined by inter-pretat; of July 1974 serial pnotographs is located 13 miles west of the Site. This distance is 5.2 times the low population zone radius of 2.5 miles.
The seg=ent within 50 miles of BFS with the largest projected population is the segment containing Tulsa, Oklahoma, which is the nest sector, between 20 and 30-mile radii. The largest projected cumulative population density area is within 30 miles of BFS, in which the city of Tulsa is located.
The seg=ent within 50 miles of BFS with the largest projected population is the segment containing Tulsa, Oklahoma, which is the nest sector, between 20 and 30-mile radii. The largest projected cumulative population density area is within 30 miles of BFS, in which the city of Tulsa is located.
Regional incorporated co== unity statistics are presented in Table 2.1-3.
Regional incorporated co== unity statistics are presented in Table 2.1-3.
Data presented are the name of the co== unity, county in which the co== unity is located, distance and direction from the Site, and the 1970 census popu-lation. Location of the above ce== unities in relation to the Site a;e shown on Figure 2.1-8.
Data presented are the name of the co== unity, county in which the co== unity is located, distance and direction from the Site, and the 1970 census popu-lation. Location of the above ce== unities in relation to the Site a;e shown on Figure 2.1-8.
2.1.3.3   Transient Population. The transient population.within a 5-mile radius of BFS central complex include school and church attendees, co==ercial and industrial employees, recreational facility employees and users, and public 2,1-4 l2S
2.1.3.3 Transient Population. The transient population.within a 5-mile radius of BFS central complex include school and church attendees, co==ercial and industrial employees, recreational facility employees and users, and public 2,1-4 l2S


e BFS TABLE 2.I-1 I'
e BFS TABLE 2.I-1 I
AREA RE510ENT POPUl.Afl0N Amo PROJECTIONS (Ref. Figure 2.1-6)
AREA RE510ENT POPUl.Afl0N Amo PROJECTIONS (Ref. Figure 2.1-6)
Redlet Distance f rom Reactor (mile) 10*lle M                  Total sector   M       M           1-1         M         ld               '4-10 25       56     22       454       363 N    1970    3            3 4                   80       31       362       513 1983    0                         36 42       94     37       425       603 1990      0           5 12m       49       570       838 2000      0           7          56 156       61       707*   1002 2010      0           8          70
Redlet Distance f rom Reactor (mile) 10*lle sector M
                                                                      ;*      857     1214 2023      0           to            64    189 8     42       310       414 NME    1973      3            0           51 0         297     8e7       60       443     1217 198)    0 0         305     501       70       $19     1395 1993    0 0         401     572       94       696     1763 2003    0 0           0        426     639     117       863   2315 1010 0         440       633     142     1%6     2268 2020      0 1973     0           8         243     674       44       222     fl91 NE 908     974     160       310     2487
M 1-1 M
      .          1983      0        135 140           932     999     191       352     2614 1990      0
ld M
(                                                                             427    3377 2003     0         I S*         1223     1311     232 1296     1389     263       496     3641 2310      0       197 0        2 38         1339     1435     293         569   3844 2020 0             5           8       33     47       210       303 Eht    1970 0             7         237 6     67       289       726 1983 0             8         298       90     79       3 14     797 1993 11         333     II*     105       363       983 2033      0 14         416     134     131       432     1097 2010      0 434       159     159       435     1207 2020      0         17 0            8         11         8     14       194       235 E    1973 0         11           16       !!      20       266       324 1933 0                       18         13     23       295       362 1993                13 0           i$           25         18     31       327       419 2000 22     39       357       471 2010      3        22            31 0         27             37       27     47       384       522 2323 11       14     0       227       260 E5s    1973      0           8 16       23       0       334       381 1933      0           11 0                        18       23       0       381     435 1990                  13 0         'S           25       !!      O       443       514 2003 0                       31       39       0       499       591 2010                  22 0                       37       47       0       551       662 2020                  27
'4-10 Total N
                              ,                2.1-21 (gj4 W,
1970 3
3 25 56 22 454 363 1983 0
4 36 80 31 362 513 1990 0
5 42 94 37 425 603 2000 0
7 56 12m 49 570 838 2010 0
8 70 156 61 707*
1002 2023 0
to 64 189 857 1214 NME 1973 3
0 51 8
42 310 414 198) 0 0
297 8e7 60 443 1217 1993 0
0 305 501 70
$19 1395 2003 0
0 401 572 94 696 1763 1010 0
0 426 639 117 863 2315 2020 0
0 440 633 142 1%6 2268 NE 1973 0
8 243 674 44 222 fl91 1983 0
135 908 974 160 310 2487 1990 0
140 932 999 191 352 2614
(
2003 0
I S*
1223 1311 232 427 3377 2310 0
197 1296 1389 263 496 3641 2020 0
2 38 1339 1435 293 569 3844 Eht 1970 0
5 8
33 47 210 303 1983 0
7 237 6
67 289 726 1993 0
8 298 90 79 3 14 797 2033 0
11 333 II*
105 363 983 2010 0
14 416 134 131 432 1097 2020 0
17 434 159 159 435 1207 E
1973 0
8 11 8
14 194 235 1933 0
11 16 20 266 324 1993 0
13 18 13 23 295 362 2000 0
i$
25 18 31 327 419 2010 3
22 31 22 39 357 471 2323 0
27 37 27 47 384 522 E5s 1973 0
8 11 14 0
227 260 1933 0
11 16 23 0
334 381 1990 0
13 18 23 0
381 435 2003 0
'S 25 O
443 514 2010 0
22 31 39 0
499 591 2020 0
27 37 47 0
551 662 2.1-21 (gj4 W,


(
(
BFS TABLE 2.5-1 (Continued)
BFS TABLE 2.5-1 (Continued)
(
(
10*lle 1-4,t  4 d        M       Total sector Ye_el
10*lle 4
_        Od       1-2     1-1 18      194      228 SE     1970     3         3         5         .
M Total d
1983     0         4         7         7       29     3I2       359 8         8       34     369       424 1990    0        5 2000     0         7       11         11       4I     442       512 0         8       14         14       48     514       598 2010 0       10       17         17       55     588       687 2020 8     17         18       0       335     381 SSE    1970    3 24         29       0       540     604 1983    0       11 0      13       28         34         0     794     869 1990 0       18       38       41         0     952   1049 2030 0       22       47         48         0   1107     1224 2010 0       27       57       55         0   1268     1407 2020 14         3         3       0     441     465 s      1970    3 0       20         4         5       0     712       74I 1953 0     23         5         6       0     S40       874 1990 2000     0       31         7         7       0   1007     1052 2010     0       39         8         8       0   1171     1226 2020     0       47       10         9       0   1340     1407 55d     1970     0         0         0       to       46       285     321 0         0         0       16     42       459     517 1933 0         0       0       19     49       541     609 1993 2000     0         0       0       23       53       %9       731 2010     0         0         0       26     69       755       850 2020     0         0         0       30     79       8%       973 0         0         0         3     10     495       508 sw    1973 0         0         0         5     16       797     818 1933 0         0         0         6     19       940     965 1993 0         0         7     23     1128     1153 2000    0 2010     0         0         0         3     26     1311     1345 2020     0         0       0         9     30     1501     1540 wsd             8         0         0         5       8     596       617 1970 0         0         0         S     13     960       981 1983 0         0         0         9     15     1124     1148 1993 2000     0         0         0       11     18     1349     1373 0         0       13     21     1563     1602
sector Ye_el Od 1-2 1-1 1-4,t SE 1970 3
  '              2010      0 2020     0         0         0       15     24     1795     1834 2.1-22
3 5
18 194 228 1983 0
4 7
7 29 3I2 359 1990 0
5 8
8 34 369 424 2000 0
7 11 11 4I 442 512 2010 0
8 14 14 48 514 598 2020 0
10 17 17 55 588 687 SSE 1970 3
8 17 18 0
335 381 1983 0
11 24 29 0
540 604 1990 0
13 28 34 0
794 869 2030 0
18 38 41 0
952 1049 2010 0
22 47 48 0
1107 1224 2020 0
27 57 55 0
1268 1407 s
1970 3
14 3
3 0
441 465 1953 0
20 4
5 0
712 74I 1990 0
23 5
6 0
S40 874 2000 0
31 7
7 0
1007 1052 2010 0
39 8
8 0
1171 1226 2020 0
47 10 9
0 1340 1407 55d 1970 0
0 0
to 46 285 321 1933 0
0 0
16 42 459 517 1993 0
0 0
19 49 541 609 2000 0
0 0
23 53
%9 731 2010 0
0 0
26 69 755 850 2020 0
0 0
30 79 8%
973 sw 1973 0
0 0
3 10 495 508 1933 0
0 0
5 16 797 818 1993 0
0 0
6 19 940 965 2000 0
0 0
7 23 1128 1153 2010 0
0 0
3 26 1311 1345 2020 0
0 0
9 30 1501 1540 wsd 1970 8
0 0
5 8
596 617 1983 0
0 0
S 13 960 981 1993 0
0 0
9 15 1124 1148 2000 0
0 0
11 18 1349 1373 2010 0
0 0
13 21 1563 1602 2020 0
0 0
15 24 1795 1834 2.1-22


r BFS TABLE 2.1-1 (Continued) 10 +11e M            5-10    Total Sector     Year   0-1     1-2       2d               4-1 8         0     0   810       821 W        1970    0      3 1983     0       5       13         0     0 1.305   I.323 0       6       IS           0     0 1.539   1.560 1990 18           0     0 1,846   1,871 2000    0       7 0      8       22           0     0 2,145   2.175 2010 0               24           0     0 2.456   2,489 2020              9 3         0     3     23       35 VK4        1970    3      3 0       5         5         0     5     37       52 1983 0       6         6         0     6     44       62 1993 2000     0       7         7         0     7     52       73 2010     0       8         8         0     8     61       85 2020     0       9         9         0     9     70       97 NW         1970     3       0       17         25   19     612       676 0       0       24         36   27     874       96l
r BFS TABLE 2.1-1 (Continued) 10 +11e Sector Year 0-1 1-2 2d M
(                  1983 1990     0       0       28         42   32   1,023   1,127 2000     0       0       38       56   43   1,374   1.511 2010     0       0       47         70     53 1,703   1,873 2020     0       0       57       84     64 2,064   2,269 NNW       1970     0       0       39       61     44     291     435 0       0       56         87   63     416     622 1983 0       0       65       102     74     487       723 1993 2000     0       0       88       137     99     653       977 2010     0       0     109       170   122     810   1,211 0     132       206   148     982   1,468 2020      0 OMNO 29     63     441       923   297   5,500   7,253 TOTALS    1970 1983     0 213     1,6)3     1,771   533   8,416   12.626 1990     0   232   1,768     1,946   629   9,997   14.572 2000     0   308   2,327     2,465   801 12,275. 18,176 0         2.525     2,706   958 14,469   21,006 20e          348 2,677     2,915 1,124 16,774   23,831 2020      0  391 2.1-23
4-1 5-10 Total W
1970 0
3 8
0 0
810 821 1983 0
5 13 0
0 1.305 I.323 1990 0
6 IS 0
0 1.539 1.560 2000 0
7 18 0
0 1,846 1,871 2010 0
8 22 0
0 2,145 2.175 2020 0
9 24 0
0 2.456 2,489 VK4 1970 3
3 3
0 3
23 35 1983 0
5 5
0 5
37 52 1993 0
6 6
0 6
44 62 2000 0
7 7
0 7
52 73 2010 0
8 8
0 8
61 85 2020 0
9 9
0 9
70 97 NW 1970 3
0 17 25 19 612 676
(
1983 0
0 24 36 27 874 96l 1990 0
0 28 42 32 1,023 1,127 2000 0
0 38 56 43 1,374 1.511 2010 0
0 47 70 53 1,703 1,873 2020 0
0 57 84 64 2,064 2,269 NNW 1970 0
0 39 61 44 291 435 1983 0
0 56 87 63 416 622 1993 0
0 65 102 74 487 723 2000 0
0 88 137 99 653 977 2010 0
0 109 170 122 810 1,211 2020 0
0 132 206 148 982 1,468 OMNO TOTALS 1970 29 63 441 923 297 5,500 7,253 1983 0
213 1,6)3 1,771 533 8,416 12.626 1990 0
232 1,768 1,946 629 9,997 14.572 2000 0
308 2,327 2,465 801 12,275.
18,176 20e 0
348 2.525 2,706 958 14,469 21,006 2020 0
391 2,677 2,915 1,124 16,774 23,831 2.1-23


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                                                                                                                                                                                                                            ~1
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                                                                            ,,--                       y                                                ,,                          __                  - . .
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.~
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~~
0           1       2 3 MILES SCALE:                 e                           i BFS 10-MILE RADIUS AP.EA HAP (REFERTOTABLE2.1-1)
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BFS 10-MILE RADIUS AP.EA HAP (REFERTOTABLE2.1-1)
F IGURE 2.1-6 2.1-48
F IGURE 2.1-6 2.1-48


9 BFS TABLE 2.1-1
9 BFS TABLE 2.1-1
(           REGIONAL INCORPORATED COMMUNITY STATISTICS Distance and           1970 County             Otractlon     Population g
(
3 miles NE               948 inola                  Roge rs Wagone r            8 miles W5W               17 New Tulse 9 miles W W               23 Fair Oaks              Wa gone r Tiewsh*               Rogers               to miles N                 95*=
REGIONAL INCORPORATED COMMUNITY STATISTICS Distance and 1970 g
Catocss               Rogers               12 miles WJ             973 Chouteau               Mayes               13 elles ENE         I.046 Coweta                 Wagone r             13 miles 55W         2.457 eroken Arrow         Tulse                 ik miles WSW       11.787 C l aremre             Rogers               14 miles Nw         9.084 Wagone r             Wa gone r             15 miles SE         4.959 Wagone r              16 miles 5             230 Red Bird Wagone r             18 miles 5             624 Porter P ryo r               Mayes                 19 miles NE         7.057 Tulse                 10 miles Wd         3,491 Owesso Tullahassee           Wa gone r             21 miles SSE             183 Muskogee             22 miles $$V         2,06)
County Otractlon Population inola Roge rs 3 miles NE 948 8 miles W5W 17 New Tulse Wagone r Fair Oaks Wa gone r 9 miles W W 23 Tiewsh*
Maskell Foyll                 Rogers               22 miles N               I64 Tulsa                 23 miles Sw         3,973 Bimby Mayes               23 miles ENE         1,090 Locust Grove Okay                   wagoner               23 miles SE             4I9 Tulse                 23 miles W       330,350 Tulsa Tulsa                24 miles Kd         3,009 Collinsville Jecks                 Tulsa                 15 miles wsw         I.997 25 mils 5               525 Taft                  Muskogee Oologeh               Rogers               25 miles h%d             458 26 miles E               82
Rogers to miles N 95*=
          # eggs                Cherokee 26 miles ENE         I,024 Salina                Mayes 26 mile s ESE           505 Hulbert                C he rckee Adair                 Meyes                 23 miles NE             459
Catocss Rogers 12 miles WJ 973 Chouteau Mayes 13 elles ENE I.046 Coweta Wagone r 13 miles 55W 2.457 eroken Arrow Tulse ik miles WSW 11.787 C l aremre Rogers 14 miles Nw 9.084 Wagone r Wa gone r 15 miles SE 4.959 16 miles 5 230 Red Bird Wagone r Porter Wagone r 18 miles 5 624 P ryo r Mayes 19 miles NE 7.057 Owesso Tulse 10 miles Wd 3,491 Tullahassee Wa gone r 21 miles SSE 183 Maskell Muskogee 22 miles $$V 2,06)
                  *Tianah is en unincorporated area within 10 miles of the plant site.
Foyll Rogers 22 miles N I64 Bimby Tulsa 23 miles Sw 3,973 Locust Grove Mayes 23 miles ENE 1,090 Okay wagoner 23 miles SE 4I9 Tulsa Tulse 23 miles W 330,350 Collinsville Tulsa 24 miles Kd 3,009 Jecks Tulsa 15 miles wsw I.997 Taft Muskogee 25 mils 5 525 Oologeh Rogers 25 miles h%d 458
# eggs Cherokee 26 miles E 82 Salina Mayes 26 miles ENE I,024 Hulbert C he rckee 26 mile s ESE 505 Adair Meyes 23 miles NE 459
*Tianah is en unincorporated area within 10 miles of the plant site.
It has been lectuoed in this I stir.g because of its proximity to the plant site.
It has been lectuoed in this I stir.g because of its proximity to the plant site.
                **Tlawah pcoulation is estimated f rom dwelling counts on the County
(
(
Hig% ay Map insert.
**Tlawah pcoulation is estimated f rom dwelling counts on the County Hig% ay Map insert.
2.1-9
2.1-9 I
* I      O
O


s -
s BFS TABLE 2.1-3 LOCAL COMMUNITY POPULATION PROJECTION & DESSITY YEAR ESTIMATED POPULATION DENSI1T Inola (3 mi. NE) 1970 948 237 1974 1,176 345 1977 2,050 512 1980 2,900 725 1983 3,080 770 1990 3,700 925 2000 4,200 1,050 2010 4,450 1,112 2020 4,600 1,150 Tiawah (10 mi. N) 1970 95 127 1974 106 141 1977 116 155 1980 125 167 1983 135 180 1990 159 212 2000 213 284 2010 264 352
BFS TABLE 2.1-3 LOCAL COMMUNITY POPULATION PROJECTION & DESSITY ESTIMATED POPULATION       DENSI1T YEAR Inola                                                             237 1970                      948 (3 mi. NE)                                 1,176                 345 1974 2,050                 512 1977 2,900                 725 1980 3,080                 770 1983 3,700                 925 1990 4,200             1,050 2000 2010                     4,450             1,112 4,600               1,150 2020 Tiawah                                                           127 (10 mi. N)         1970                       95 106               141 1974 1977                     116                 155 1980                       125               167 1983                       135               180 1990                       159               212 2000                       213               284 2010                       264               352 321               428
(
(                          2020 Residents per square mile.
2020 321 428
*Residents per square mile.
(
(
2.1-15
2.1-15


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Latest revision as of 15:15, 4 January 2025

Responds to NUREG-0578, TMI-2 Lessons Learned Task Force Status Rept & Short-Term Recommendations, IE Bulletin 79-08 & Selected Emergency Preparedness Issues.Reiterates 790615 Analyses in New Format.Certificate of Svc Encl
ML19225C628
Person / Time
Site: Black Fox
Issue date: 07/27/1979
From: Ewing T
PUBLIC SERVICE CO. OF OKLAHOMA
To: Varga S
Office of Nuclear Reactor Regulation
References
6212.125.3500.2, NUDOCS 7908010573
Download: ML19225C628 (64)
v • d • e


Text

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6212 DIN 8-016-854 PUBUC S%1CE COMPANY OF OKLAHOMA V '}- q I

A CENTRAL AND SOUTH WEST COMPANY f

a P.O. box 201/ TULSA. OKLAHOMA 74102 / (918) 583-3811 L

s c

Public Service Company of Oklahoma July 27, 1979 Black Fox Station File: 6212.125.3500.21L Response to lessons Learned Report 6212.217.0521.21L USNRC Docket Nos. STN 50-556, 50-557 Mr. Steven A. Varga, Assistant Director Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20002

Dear Mr. Varga:

The meeting between our respective organizations on July 19 in Bethesda was very productive. We especially appreciated your positive comments concerning our analysis of the lessons to be learned from TMI-2 as they apply to the Construction Permit application for the Black Fox Station. This analysis, which was sent to Mr. Harold R.

Denton, Director, Nuclear Reaction Regulation, on June 15, 1979, represented the initial effort by Public Service Company of Oklahoma (PS0) to respond to the events at TMI and documented our long-term corporate commitment to fully analyze every facet of the TMI-2 accident arid to incorporate the lessons learned into the design, construction, staffing, training and operation of the Black Fox Station.

With the issuance on July 19 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," e compared the 23 lessons learned in that report with the PSO analysis. Our understanding of NUREG-0578 and the comparison of the two documents were greatly facilitated by the helpful explanations and advice offered by you and your staff during our meeting on July 19. These discussions, along with inforration provided by Mr. Denton during his meeting with our President, Mr. R. O. Newman on July 20, enable us to respond promptly to your request for com-mitments to the requirements and recommendations of NUREG-0578.

Although our June 15 lessons learned analysis adcrassed most of the issues discussed in NUREG-0578, the organization of the material is different. Consequently, to facilitate your review, we are reiterating our concitments in a format consistent with the organization of NUREG-0578.

In addition to specifically addressing every recommendation and requirement of NUREG-0578, this submittal also addresses matters applicable to Black Fox which were developed by the Bulletins and Orders Task Force, and the Emergency Preparedness group headed by Mr. Brian Grimes.

ll 4 I 8 O (o 3 k'

68 cqG h

7908010 "..

- CENTRAL AND SOUTH WEST SYSTEM

~

T Sc ge n E egnc Power ges4,egsystes

% ' Qe7a!?ger.gd L:ght gcjefv ce Company of CManoma

a Mr. Steven A. Varga, Assistant Director Page 2.

We concur with the view presented during the meetings of July 19 and 20, that all of the commitments and actions raquired of us 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 for the Black Fox Station. Our commitments reflect this understanding and philosophy.

The TMI-2 accident has stalled progress on the Black Fox application, and as you know, we are quite anxious to overcome this licensing delay.

Consequently, we have responded directly and completely to all of the issues applicable to the Black Fox Station as presented by the two Task Forces and Mr. Grimes's group; this submittal should satisfy all of those concerns.

In these circumstances, we do believe it reasonable to expect the NRC Staff to complete its report quickly and to respond to the Licensing Board Orderof June 13, 1979 in the very near future.

Please call Mr. Vaughn Conrad, Manager, Licensing and Compliance at (918) 583-3611 if you have any questions regarding this submittal.

Sincerely yours, b[ _.

T. N. Ewing, Manager Black Fox Stationplear Project TNE:VLC:dm Attachment xc:

(w/ attachment) BFS Service List (o '{

BLACK FOX STATION SERVICE LIST CERTIFICATE OF SERVICE I hereby certify that a copy of the foregoing PS0 Response to the TMI Event has been served on each of the following persons by deposit in the United States mail, first-class postage prepaid, this 27th day of July, 1979.

L. Dow Davis, Esquire Mr. Joseph Gallo Counsel for NRC Staff Isham, Lincoln & Beale U. S. Nuclear Regulatory Comission 105017th Street N. W.

Washington, D. C.

20555 Washington, D. C.

20036 Mr. Cecil 0. Thomas Joseph R. Farris, Esquire U. S. Nuclear Regulatory Cormiission Green, Feldman, Hall & Woodard Phillips Building 816 Enterprise Building 7920 Norfolk Avenue Tulsa, Oklahoma 74103 Bethesda, Maryland 20014 Docketing and Service Section Andrew T. Dalton, Esquire Office of the Secretary of the Comn.

1437 South Rin Street, Suite 302 U. S. Nuclear Regulatory Comission Tulsa, Oklahoma 74119 Washington, D. C.

20555 (20 copies)

Mr. William G. Hubacek Mrs. Ilene H. Youngnein U. S. Nuclear Regulatory Commission 3900 Cashion Place Office of Inspection and Enforcement Oklahoma City, Oklahoma 73112 Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76012 Mr. Gerald F. Diddle Mr. Lawrence Burrell General Manager Route L, Box 197 Associated Electric Cooperative, Inc.

Fairview, Oklahoma 73737 P. 0. Box 754 Springfield, Missouri 65801 Mr. Maynard Human Mrs. Carrie Dickerson General Manager Citizens' Action for Safe Energy, Inc.

Western Famers Electric Ccoperative P. O. Box 924 P. O. Box 429 Claremore, Oklahoma 74017 Anadarko, Oklahoma 73005 Michael I. Miller, Esq.

Charles S. Rogers, Esq.

Isham, Lincoln & Beale Assistant Attorney General Cne 1st National Plaza 112 State Capitol Building Suite 4200 Oklahoma City, Oklat ma, 7 105 g

Chicago, Illinois 60603 g

)

i OA%%rs iW Manager,LicensinN(\\

Vaughn L.qonr3t!

nd Compliance

[g 5 Public Service Company of Oklahoma

RESPONSE OF PUBLIC SERVICE COMPANY OF OKLAHOMA BLACK FOX STATI0f:, UNITS 1 & 2 USNRC DOCKET N05. 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 Issues on Emergency Preparedness July 27, 1979.

hg

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-0perated Relief Valves and Block Valves and Pressurizer Level Indicators in PWR's.................. 4 2.1.2 Performance Testing for BWR and PWR Relief and Safety Va l v es...........................

5 2.1.3.a Direct Indication of Power-0perated Relief Valves and Safet3 Valve Position for PWR's and BWR's.............

6 2.1.3.b Instrumntation for Detection of Inadequate Core Cooling in PWR's and BWR's....................... 7 2.1.4 Diverse and More Selective Containment Isolation Prcvisions for PWR's and BWR's....................... 9 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident P u rg e Sys tems........................ 10 2.1.5.b Inerti ng BWR Containments.................... 11 2.1.5.c Capability to Install Hydrcgen Recombiner at Each Light Water Nuclear Power Plant..................... 12 2.1.6.a Integrity of Systems Ottside 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.s.b Auxiliary Feedwater Flow Indication to Steam Generators for PWR's............................

16 1.1.8.a Improved Post-Accident Sampling Capability...........

17 2.1.8.3 Increased Range of Radiation Monitors.............

.18

Table of Contents Page 2.1.8.c Improved In-Plant Iodine Instrumentation...........

19 2.1.9 Analysis of Design and Off-Normal Transients and Accidents........................

20 2.2.1.a Shift Supervisor's Responsibilities.............

23 2.2.1 b Shift Technical Advisor...................

25 2.2.1.c Shift and Relief Turruver 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 Safaty Syrtem nvailability.......

30 Response to Inspection and Enforcem:nt Bulletin 79-08...........

32 Response to Selected Issucs on Emergency Preparedness...........

45 0

INTRODUCTION AND DESCRIPTION OF METHODOLOGY On June 15, 1979, PuMic 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 tL 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 Committee 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, h,

For the balance, the intent of each commitment by P50 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 neeting 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 Bulletiri (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 commitment 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 cperation.

The PS0 commitnents to action require corpletion 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 scme duplication of subject matter with the possibility of different interpretations of the PSO response between the two task forces.

If such differences are identified, PSO commits to work with the NRC Staff to reconcile them.

-;L'

~

(

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 PS0 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

Therein, of the State of Oklahoma, George Nigh to Joseph Hendrie, Chairman USNRC.

the status of the State Emergency Response Plan, PS0's role in development, and a connitment to have a NRC approved plan in effect well before BFS conmercial 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 actions required by the NRC Staff can be satisfied during the post-construction permit phase of the Black Fox design and construction effort, and that th documentation of these activities should be set forth in the Final Safety Analysis Report and Station Operating Procedures for the Black Fox Station. Our co=nittents reflect this understanding and philosophy.

RESPONSE TO NUREG-0578, Appendix A TMI-2 Lessons Learned Task Force Short-Term Recoundations 72

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Emergency Power Supply Reauf rements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWR'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.

4_

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Performance Testina 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 maximized. 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 d

PS0 believes that it is important to assure that the safety and relief valves installed in the BFS reactor coolant bnundary will function as intended and maintain their integrity under exoected 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 necesssry testing to qualify the reactor coolant system relief and safety valves for the ecst 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 applicaticn.

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Direct Indication of Power-Ocerated Relief Valve and Safety Valve Position for PWR's and BWR's Section 2.1.3.a NRC STAFF POSITION Reactor system relief and safety valves shall be provided with a positive indi-cation in the control room derived from a reliable valve position detection 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 F5AR. 75

NRR Lessons Learned Task Force Short-Term Reco mendations TITLE:

Instrumentation for Detection of Inadeouate Core Cooling in PWR's and BWR's (Section 2.1.3.b NRC STAFF POSITION 1.

Licensees shall develop procedures 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 descriptirn of the analyses needed to form the basis for operator training and procedure development shall be provided,1ursuant to another short-term requirement, " Analysis of Off-Nomal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).

In addition, each PWR shall install a primary coolant saturation meter 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 of the functional design requirements for the system shall also be included.

A descripticn 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 unambiguously deternine 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 cperator in recognizing inadequata core cooling, and oparators will be throroughly trained in the procedure and utilization of instrumentation to assure correct interpretation of the core h

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.

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 recontendations 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 accordingly, and shall report the results of the re-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 automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of con-tainmentisolation valves shall require deliberate operator action.

PS0 COMMITMENT P50 recognizes the importance for timely and effective isolation of the containment under accident conditions. P50 will review the design of BFS to assure that the final design provides for:

1.

Diversity in the parameters sensed for the initiation of containment isolation, in accordance with SRP 6.2.4; 2.

Automatic isolation of non-essential systems upon containmer,t 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 information on the definition of essential and non-essential systems. 78

NRR Lessons learned Task Force Short-Term Recommendations TITLE: Dedicated Penetrations for External Recombiners or Post-Accident Purce 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 containment of each unit. This position is therefore not applicable.

NRR Lessons learned Task Force Short-Term Recommendations TITLE: hertina BWR Containments (Section 2.1.5.b).

NRC STAFF POSITION It shall be required that the Vemont 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. /b,0

NRR Lessons Learned Task Force Short-Term Recommendations TITLE:

Capability to Install Hydrogen Recombiner at Each Light Water Nuclear 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 an 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 T?il-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. l

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Integrity of Systems Outside Containment likely to Contain Radioactive Materials (Encineered 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 e program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as practical levels. This program shall include the following:

1.

Imediate Leak Reduction.

Implement all practical leak reduction measures for all systems a.

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.

PS0 COF"41TMENT 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 oreventive 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 wil1 contain the results of the above desig" and ope.ations review.

, g 2-

NRR Lessons Learned Task Force Shcrt-Term Recorrendations 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-accide'it 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 ue unduly limited or safety equipment may be 0.' duly degr.ided by the 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.

P50 COMMITMDE PS9 recognizes, as a result of the TMI-2 event, the need to assure necessary access to vital areas and protecticn 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 described in Regulatory Guide 1.3, will evaluate the BFS design for unacceptable limitations on personnel access and occupancy or undu" degradation of 2fety-related equipment curing post-acciocct operations. Thr evaluation will consider alternatives, including layout changes, increased use of permanent shielding, temporary shielding, or proce.iural controls.

The The evaluation will determine changes needed throughout Black Fox Station.

results of the evaluation and a description of the changes will be reflected in the final design presented in the FSAR. o

NRR Lessons Learned Task Force Short-Term Recorrendations TITLE: Automatic Initiation of the Auxiliary Feedwater System for PWR's (Section 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.

'5-lb o

NRR Lessons Learned Task Force Short-Tern Recorrendations TITLE: Auxiliary Feedwater Flow Indication to Steam Generators for PWR's Section 2.1.7.b This issue is not applicable to the BWR/6 Nuclear Steam Supply System of the Black Fox Station, Units 1 and 2.

NRR Lessons Learned Task Force Short-Term Recommendations TITLE:

Improved Post-Accident Samplino 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 detemine 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 spectrun analysis facilities shall be performed to determine the capability to promptly (less than 2 hcurs) 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 reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.

The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.

If the review indicates that the analyses required cannot be perforemd in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

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

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

NRR Lessons Learned Task Force Short-Tem Recomendations TITLE:

Increased Range of Radiation Monitors (Section 2.1.8.b).

NRC STAFF p0SITION The requirements associated with this recommendation should be considered as advanced implementation of certain requiret.2nts 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 oeprating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.

5 a.

Noble gas effluent monitors with an upper range capacity of 10 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 uti/cc (Xe-133) to a maximum of 105 uCi/cc (Xe-133). Multiple monitors are considered to be necessary 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 provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis.

8 3.

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

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

7 NRR Lessons Learned Task Force Short-Term Reco= endations 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 COMMITPENT PS0 will provide instrumentation, training of personnel and the technical procedures for accurately determining airborne iodine concentration throughout the plant ur.de" accident conditions, with documentation to be provided in the FSAR. 86

NRR Lessons Learned Task Force Short-Term Recomendations TITLE: Analysis of Desien 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-coolant 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 Tap 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 emergency proce-dures and in support of an elentual long-term verification of cogliance 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 forced 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 approached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real tire 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 coerators 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 to 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 present, 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 ccmplete loss of auxiliary feedwater 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 shr.11 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 infomation 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 more serious accidents.

The information derived 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 trane!ents should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for ccmparisons 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 rathods 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 operaticn of the Black Fox Station.

PS0 understands that analysis and emergency procedures or guidelines for:

1.

Smdl break loss-of-coolant accidents; 2.

Inadequate core cooling; and C)0 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 0 der Task Force. These analyses are being generalized first to cover BWR/1-5 type power plants and will be extended by General Electric Copany to cover the BWR/6 System generically. Each of the specific requirements stated in the above position tue been identified by the Bulletins and Orders Task Force. As this asscssnent 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 amergency procedures guidelines, and these guidelines will be implemented in the Black Fox Station procedures and training programs as appropriate. c,

NRR Lessons Le:rned Task Force Short-Term Recommendations TITLE: Shift Supervisor's Responsibilities (Section 2.2.1.a).

NRC STAFF POSITION 1.

The highest level of corporate management of each licensee shall 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 room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command 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 on duty in the control room. The idea shall be reinforced that 1he shift supervisor should not become totally involved in any single operation in times of emergency when mLltiple 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 he activities of control rtom operators.

Persons authorized to rt ieve the shift supervisor shall be specified, c.

If the shift supervisor is temporar.ily absent from the control room during routine operatior.3, a lead control room 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 Jtility responsible for plant operations. Admini-strative functions that detract from or are subordinate to the management responsibility for assu ing the safe operation mf the plant shall be delegated to other operations personnel not on duty in the control room. c4

Shif t Supervisor's P.esponsibilities (Section 2.2.la)--

Continued.

pSO COMMITMENT PS0 comits 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 plant 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. g3

NRR Lessons Learned Task Force Short-Term Recommendations 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 shall have a bachelor's degree or equivalent in a scientific or engineering discipline and 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.

MRC 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.

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 complete and sign.

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

Assurance that critical plant parameters are within allowable a.

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 anr' =idents by a check of the control console (what to check criteria for acceptable status shall be included on the checklist);

ano Identification of systems and components that are in a degraded c.

mode of operation permitted 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 separa+e 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).

PSO COMMITENT PS0 commits to compliance with the above position and concurs that it is a prudent management approach to plant operations. 95

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 roem operators),

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 administrative 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 authority and responsibility in the control room in the event af an emergency. The line of succession for the person in charge of tre control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall claarly 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 PSO will comply fully with this position and recognizes the importance of access control to the control room. N/)D

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

NRC STAFF POSITION Each operating nuclear power plant shall raintain 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 managemer.t support of reactor operations in the eveat of an accident. The center shall be habitable to the same degree as the centrol room for postulated accident conditions.

The licensee shall revise his emergency plans as necessary to incorporate the role and location of the tichnical 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 6ccessible to the technical support center under emergency conditions. These documents shall incluce, 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 COMMITMENT An onsite technical support center as described above will bc

-sith the capability to display necessary plant status information for 1,i is who are knoveledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to 'he 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. 9'7

NRR Lessens Learned _ Task Force Short-Term Recommendations TITLE: Onsite Operational Support Center (Section 2.2.2.c).

NRC STAFF POSIT 7p fn 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 o which the operations support personnel will report in an emergency situation.

omunications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of comunication and ranagement.

PS0 C0KMTMENT PS0 will designate an area to serve as the operational support center as described in the above position. The support canter will be physically separated from the control room, and appropriate comun1 cation 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 Suppcrt Center, as well as establish the methods and lines of communication and management control. The location of the Center will be provided in the FSAR. gg

NRR Lessons Learned Task Force Short-Term' Recommendations TIi ': 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 threshold of complete loss of safety function.

Identification of a huran or operational error that prevents or could prevent the ai complishment of a safety function required by NRC regulations and analy7ej in the license application shall require placement of the plant in a hot sht.tdown ccidition 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 baurs.

The loss of operability of a saft-;v function shall include consideration of the ncessary instrumentation, cetrols, 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 measures 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.

)h 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.

PS0 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 for under the Administrative Procedures Act. This approach was emphasized in Dr. Mattson's letter of July 18,1979 to Mr. Denton, attav During the July 20 meeting with PS0, 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.50 at this tire. PS0 does agree to comply with any requirement ultimately determined by the rulemaking.

. [60

ga ascg UMTED STATES 6,.

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3,,(g g NUCLEAR REGUL ATORY COMMISSION W ASHINGTON D. C. 20555 3g 7

QQ.f July 18,1979 MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:

Roger J. Mattson, Director THI-2 Lessons Learned Task Force TMI-2 LESSONS LEAR"ED TASK FORCE

SUBJECT:

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 There are construction, and pending construction permit appliutions.

23 specific recomenc'ations 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 reco=endations have a majority concurrence by the The excepi. ion is the recommended requirement to provide Task Force.

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

Three of the recomendations appear to require changes in existing regulations for which the Task Force recommends immediately effective

1) inerting of MKI and MK II BWR containments that rulenaking. They are:

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 er operational error.

to develop the required Commission papers and carry through with these rulemaking actions.

The 23 recomended actions were discussed with the Regulatory Requirements Review Committee (June 22, 1979), the Commission (June 25,1979), the 11,1979), and the ACRS (July 12, 1979).

TMI-2 Subcommittee of the ACRS (July 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.

l O(

2 Harold R. Denton The Task Force recommends that time not be taken to request and evaluate public comments on these short tem requirements prior to their promulgation as licensing requirements or rules because they are safety significant matters that require prompt application to operating reactors and operating license Other Tl1I-2 accident review groups applications in the late stages of review.

and the Lessons Learned Task Force are continuing to evaluate the longer term implications of the accident. Any public coments on the short tem 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 tem 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 span of the These longer Task Force) before other regulatory actions are recommended.

term interests of the Task Force are described in Section Three of the The Task Force intends to develop its final recommendations and report.

issue a final report in early September 1979. The topics to be addressed 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 11, 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 tem reco=endations 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 by the various NRR organizations in the course of their review. Where appropriate, we made clarifying changes in the language of the report. The principal substantive change occurred in the fom and schedules of the implementation section (Appendix B). Some of the com ents addressed matters that the Task There are Force has deferred for consideration in its final report.

significant differences of opinion within the staff on two of the Task Force a) the need for recommendation 2.2.3 concerning reco=endations, 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) and b) the need for the minority Task Force recommendation 2.1.5.c concerning rulemaking for backfit of recombiner capability (some support the minority recommendation, others do Having considered these co=ents and made changes to the report where not).

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

direct the i=ediate implementation by DPM, DOR or B&OTF, as appropriate, of all the short tem recommendations, except the three rulemaking a.

matters, through the issuance of licensing positions to operating plant licensees, plants under construction, and construction permit applicants.

( 0 2-

3 Harold R. Denton request the fomulation of immediately effective rules by the Office of Standards Development for action by the Comission on the three b.

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 RogerJ.dattson,' Director TMI-2 Lessons Learned Task Force

Enclosure:

as stated cc: Chaiman Hendrie Comissioner Gilinsky Commissioner Kennedy Comissioner Bradford Comissioner Ahearne ACRS (20)

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

C. Kamerer, CA (20) tiRC PDR

'O3

RESPONSE TO INSPECTION & ENFORCEMENT BULLETIN 79-08.

O L'-

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.

This review should be directed toward understanding:

(1) the extreme a.

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)(3) the necessity the apparent operational errors which led to eventual core damage; and to systematically analyn plant conditions and parameters and take appropriate corrective action; Operational personnel should be instructed to:

(1) not override autoratic b.

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 indicaticn when one or more confirratory indications are available; All licensed operators and Plant management and supervisors with operational c.

responsibilities shall participate in this review and such participation shall be documntedin plant records.

PS0 C0??iITMENT public Service Company of Oklahoma has established a Technical Advisory Comittee (TAC) to assess the events at Three Mile Island, Unit 2, and to apply the lessons learned to its Black Fox Station Project. This comittee was established at the direction of the President and Chief Executive Officer of the Company and reports its findings and recomendations directly to the Review and Audit Ccmittee.

These findings and recc=endations 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 "Lesscns Learned Task Force"--NUREG-0578 The President's Cc= mission on Three Mile Island EPRI--Nuclear Sa fety Analysis Center

[f IEB 79-08 Task 1--Continued.

Generic vendor programs Atomic Industrial Forum TMI Policy Committee NRC Special Investigation (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 recomencations from them.

The assessment 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 P50 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 precedures, staffing and training progrem, 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 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--Continvod.

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 available.

(See also commitments made under ICB 79-08 Task 5).

The Manager, Black Fox Station 4 the Manager, Nuclear Training are assigned to the TAC to ensure that operationa. 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 tc review administrative mechanisms to ensure that lessons learned are incorporated into the station training programs.

Findings and recorr.endations from the TAC will be documented in the Project files and conformance with each specified commitment will be incorporated into th c documentation system. (o9

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 ranual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.

PSO 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 cr ranual, and any necessary ranual actions will be covered by apr;opriate procedures. l06

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 tranual action necessary, describe in sunnary form the procedure by which this action is, taken in a timely sense.

PS0 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 systers that are used when the rain feedwater system is not operable will be reviewed. Both automatic and ranual actions will be assessed for adequacy, and any necessary manual actions will be addressed by procedures to assure timely actuations.

. (09

IEB 79-08 Task 4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems. Describe other redundant instrumentation which the operator might have to give the same infomation regarding plant status.

Instruct operators to utilize other available information to initiate safety systems.

p50 COM'11TMENT 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 automatic and manual initiation of safety systems will be reviewed.

Redundant instrumentation which the operator will have to give the same vessel level indications will be identified and factored into operator training, instruction, and procedures. lIO

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

Operators do not override automatic actions of engineered safety a.

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 vcssel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.

P50 COMMITMENT Prior to completion of operatin; procedures and training instructions, actions directed by tt.:se ir.structions will be reviewed to ensure that:

Operators are directed not to override automatic action of engineered a.

safety features enless 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 acticns, but to also examine other plant parameter indications in evaluating plant conditiens. fll

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 featur?s. 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 follcwing necessary ranipulations and are raintained in their proper positions dt ring all operational modes.

pSO COMMITMENT At the time of final design, i.e., FSAR submittal, PS0 will review ali safety-related valve positioning requirements and positive cont ols 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 procedures for raintenance, 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 ranipulations and are maintained in the proper position during all operational modes. ( (1

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 sur.h 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 assured.

pS0 COM'4ITMENT 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 other 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 -

ff3

IEB 79-08 Task 8 Review and modify as necessary your maintenance and test procedures to er.sure 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.

PSO 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 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 or returned to service. ik

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 connunication channel shall be established and maintained with NRC.

PS0 COMMITMENT Prior to completion of the emergency plan and implementing procedures, NRC notification shall be incorporated 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 the time of NRC notification, an open continuous connunication channel will be established and maintained with NRC. fl5

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 time 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.

43-l

IEB 79-08 Task 11 Propose changes, as required, to those technical specifications 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. II7

RESPONSE TO SELECTED ISSUES ON EMERGENCY PREPAREDNESS t

r

Emergency Preparedness 1.

Reculatory Guide 1.101 Emergency Planning 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 regulatory staff have indicated i. hat revisions to the uniform action level criteria will be forthcoming as a result of the experiences at TMI. PS0 will utilize these criteria in development of the BFS Emergency Plan.

ii.

Improved Samoling and Instrumentation Cacability.

These issues are covered in NUREG-0578 TMi-2 Lessons Learned Task Force Status Recort and Short-Term Recommendations as issue 2.1.8.

PS0 has addressed these requirements in our response to that section.

iii. Emercency Operatina 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,

'e plant status and meteorological and federal authorities involved. Appros 6ta will be read directly from instrumentation ) laced in the center.

iv.

Imoroved 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 Emerger;cy Perparedness - iv.

(Continued).

possible use of continuous radiation monitors with remote readout.

PS0 also ccmits 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 Protecti te 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 Dapartment 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 Comission, 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 and submittal to NRC approval.

Concurrently, PS0 is establishing target tasks for the BFS Emergency Response Plan development. The plan will be submitted with the CSAR in support of the application for operating licenses.

Our understanding from recent discussions 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 4,00u meters as reflected in the BFS Preliminary Safety Analysis Report and Environmental Report.

3ccordingly, we have reviewed the applicable discussion from the ER (@ 2.1.3.1) on the popluation projections within a ten-mile radius of the site. Also studied were PSAR tabluations of regional incorporated community statistics and population

-4E-

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 tiew Tulsa (eight miles WSW),

Fair Oaks (nine miles WiiW), 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 ten-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 uniformly divided into square mile sections, each of the perpendicular lines forming uniform squares on the figure represents a transportation route.

As a result of our review, we have ccncluded that implementation of protective measuras such'as evacuation is feasible over the lifetime of the station based on population estimates and evacuation routes.

vi.

Periadic Testina.

that PS0 comments to periodically corduct local emergency plan testing to assure [7d

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.

e

-4 3-ILZ

STATE OF OKL AHOM A

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7 ".0 *1 OFFIG OF THE GOVERNOR g/ J 212 STATE CAPITOL BUILDING 1'

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OKLAHOMA CITY. OKLAHOMA 73105 GEORGE NIGH June 2,0, 1979 45 / $21234.5 Mr. Joseph M. Hendrie, Chairman U. S. fluclear Regulatory Comission Washirgton, D. C.

20555

Dear Mr. Hendrie:

I share your concern with regard to states having ade-quate radiological emergency response plans in opera-tion which support fixed nuclear facilities.

I appreciate your kind offer to assist in preparing such a plan through the mechanism of the Federal Interagency Re-gional Advisory Connittee and your agency.

The Occupational r nd Radiological Health Service of the Oklahoma Depa.tment of Health, in cooperation with the Oklahoma Office of Civil Defense, has recently completed a preliminary draft of Oklahcma's radiologi-cal plan. Copies of this draft have been circulated to my office, several State executive agencies, the tiRC Office of State Programs, and Public Service Company of Oklahoma for comments. Following revision in accord with these comments, the plan will be circulated for comment to these State agencies,' local officials, the public, and the tiRC. Our current schedule calls for a final version of the plan to be ready by early 1980.

He fully intend and expect to receive iiRC concurrence to the final plan several years prior to the now anticipated operational status of the Black Fox Station in 1985.

interel r yours, c

U George ligh t

e,

BFS 2.1.3.1 Population Within 10 Miles. A map of the 10-mile area of the BFS

(

Site is presented on Figure 2.1-6.

The map is overlayed with concentric circles, centered on the central plant complex with radii of 1, 2, 3, 4, 5, I

and 10 =11es, and with radial lines forming 22-1/2 degree sectors centered on the 16 cardinal conpass points. Table 2.1-1 presents the corresponding pro-jected residential population within each annular and radial sector segnents for the expected first year of plant operation (1983) and by census decade beginning with 1990 through the end of the anticipated plant life (2020).

The largest cu=ulative population density for this area through the year 2020, occurs within the 4-=ile 1.dius area, in which the town of luola is located.

The 10-mile radius area is pri=arily rural and is expected to re=ain as such during the period of plant operation. Base data and methodology of population projections are presented in Subsection 6.1.4.2.

(

Supplement 6 2.1

  • 120376 t

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BFS The town of inola is the only significant population concentration with-(

The IQ74 esti=ated population of Inola is 1176 with in the 10-mile area.

projections presented in the Co== unity Development Plan, Inola Oklahe=a, in-creasing to 4200 by the year 2000 (6). There are three other small co== unities wA:hin 10 miles of BFS in addition to the town of Inola. The other co== unities are New Tulsa (8 miles WSW), Fair Oaks (9 miles WNW), and Tiawah (10 miles N).

New Tulsa and Fair Oaks are incorporated entities in Wagoner County while Tiawah is unincorporated and located in Rogers County. New Tulsa and Fair Oaks populations are not expected to increase significantly according to projections by the Ohiaho=a Employ =ent Security Co==ission (7). Much of the Tiawah current, estimated population of 95, is located beyond the 10-mile radius (8).

2.1.3.2 Pooulation Jetween 10 and 50 Miles.

Figure 2.1-7 shows the region within 50 miles of the reactor locations in northeast Oklahoma with concentric circles drawn at 10-mile radius intervals and wi:S radial lines defining sectors centered on the 16 cardinal compass firections. The projected pcpu-lations for 1983,1990, 2000, 2010, and 202C for each annular and radial sector segments are presented in Table 2.1-2.

The =ethods for estimating population distribution are descri'aed in subsection 6.1.4.2.

The nearest population center (as defined in 10 CFR 100) at the tire of startup of Unit 1 is Tulsa, Oklahoma with a 1970 census population of 330,350 (9). The nearer boundary of the densely populated area of Tulsa deter =ined by inter-pretat; of July 1974 serial pnotographs is located 13 miles west of the Site. This distance is 5.2 times the low population zone radius of 2.5 miles.

The seg=ent within 50 miles of BFS with the largest projected population is the segment containing Tulsa, Oklahoma, which is the nest sector, between 20 and 30-mile radii. The largest projected cumulative population density area is within 30 miles of BFS, in which the city of Tulsa is located.

Regional incorporated co== unity statistics are presented in Table 2.1-3.

Data presented are the name of the co== unity, county in which the co== unity is located, distance and direction from the Site, and the 1970 census popu-lation. Location of the above ce== unities in relation to the Site a;e shown on Figure 2.1-8.

2.1.3.3 Transient Population. The transient population.within a 5-mile radius of BFS central complex include school and church attendees, co==ercial and industrial employees, recreational facility employees and users, and public 2,1-4 l2S

e BFS TABLE 2.I-1 I

AREA RE510ENT POPUl.Afl0N Amo PROJECTIONS (Ref. Figure 2.1-6)

Redlet Distance f rom Reactor (mile) 10*lle sector M

M 1-1 M

ld M

'4-10 Total N

1970 3

3 25 56 22 454 363 1983 0

4 36 80 31 362 513 1990 0

5 42 94 37 425 603 2000 0

7 56 12m 49 570 838 2010 0

8 70 156 61 707*

1002 2023 0

to 64 189 857 1214 NME 1973 3

0 51 8

42 310 414 198) 0 0

297 8e7 60 443 1217 1993 0

0 305 501 70

$19 1395 2003 0

0 401 572 94 696 1763 1010 0

0 426 639 117 863 2315 2020 0

0 440 633 142 1%6 2268 NE 1973 0

8 243 674 44 222 fl91 1983 0

135 908 974 160 310 2487 1990 0

140 932 999 191 352 2614

(

2003 0

I S*

1223 1311 232 427 3377 2310 0

197 1296 1389 263 496 3641 2020 0

2 38 1339 1435 293 569 3844 Eht 1970 0

5 8

33 47 210 303 1983 0

7 237 6

67 289 726 1993 0

8 298 90 79 3 14 797 2033 0

11 333 II*

105 363 983 2010 0

14 416 134 131 432 1097 2020 0

17 434 159 159 435 1207 E

1973 0

8 11 8

14 194 235 1933 0

11 16 20 266 324 1993 0

13 18 13 23 295 362 2000 0

i$

25 18 31 327 419 2010 3

22 31 22 39 357 471 2323 0

27 37 27 47 384 522 E5s 1973 0

8 11 14 0

227 260 1933 0

11 16 23 0

334 381 1990 0

13 18 23 0

381 435 2003 0

'S 25 O

443 514 2010 0

22 31 39 0

499 591 2020 0

27 37 47 0

551 662 2.1-21 (gj4 W,

(

BFS TABLE 2.5-1 (Continued)

(

10*lle 4

M Total d

sector Ye_el Od 1-2 1-1 1-4,t SE 1970 3

3 5

18 194 228 1983 0

4 7

7 29 3I2 359 1990 0

5 8

8 34 369 424 2000 0

7 11 11 4I 442 512 2010 0

8 14 14 48 514 598 2020 0

10 17 17 55 588 687 SSE 1970 3

8 17 18 0

335 381 1983 0

11 24 29 0

540 604 1990 0

13 28 34 0

794 869 2030 0

18 38 41 0

952 1049 2010 0

22 47 48 0

1107 1224 2020 0

27 57 55 0

1268 1407 s

1970 3

14 3

3 0

441 465 1953 0

20 4

5 0

712 74I 1990 0

23 5

6 0

S40 874 2000 0

31 7

7 0

1007 1052 2010 0

39 8

8 0

1171 1226 2020 0

47 10 9

0 1340 1407 55d 1970 0

0 0

to 46 285 321 1933 0

0 0

16 42 459 517 1993 0

0 0

19 49 541 609 2000 0

0 0

23 53

%9 731 2010 0

0 0

26 69 755 850 2020 0

0 0

30 79 8%

973 sw 1973 0

0 0

3 10 495 508 1933 0

0 0

5 16 797 818 1993 0

0 0

6 19 940 965 2000 0

0 0

7 23 1128 1153 2010 0

0 0

3 26 1311 1345 2020 0

0 0

9 30 1501 1540 wsd 1970 8

0 0

5 8

596 617 1983 0

0 0

S 13 960 981 1993 0

0 0

9 15 1124 1148 2000 0

0 0

11 18 1349 1373 2010 0

0 0

13 21 1563 1602 2020 0

0 0

15 24 1795 1834 2.1-22

r BFS TABLE 2.1-1 (Continued) 10 +11e Sector Year 0-1 1-2 2d M

4-1 5-10 Total W

1970 0

3 8

0 0

810 821 1983 0

5 13 0

0 1.305 I.323 1990 0

6 IS 0

0 1.539 1.560 2000 0

7 18 0

0 1,846 1,871 2010 0

8 22 0

0 2,145 2.175 2020 0

9 24 0

0 2.456 2,489 VK4 1970 3

3 3

0 3

23 35 1983 0

5 5

0 5

37 52 1993 0

6 6

0 6

44 62 2000 0

7 7

0 7

52 73 2010 0

8 8

0 8

61 85 2020 0

9 9

0 9

70 97 NW 1970 3

0 17 25 19 612 676

(

1983 0

0 24 36 27 874 96l 1990 0

0 28 42 32 1,023 1,127 2000 0

0 38 56 43 1,374 1.511 2010 0

0 47 70 53 1,703 1,873 2020 0

0 57 84 64 2,064 2,269 NNW 1970 0

0 39 61 44 291 435 1983 0

0 56 87 63 416 622 1993 0

0 65 102 74 487 723 2000 0

0 88 137 99 653 977 2010 0

0 109 170 122 810 1,211 2020 0

0 132 206 148 982 1,468 OMNO TOTALS 1970 29 63 441 923 297 5,500 7,253 1983 0

213 1,6)3 1,771 533 8,416 12.626 1990 0

232 1,768 1,946 629 9,997 14.572 2000 0

308 2,327 2,465 801 12,275.

18,176 20e 0

348 2.525 2,706 958 14,469 21,006 2020 0

391 2,677 2,915 1,124 16,774 23,831 2.1-23

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BFS 10-MILE RADIUS AP.EA HAP (REFERTOTABLE2.1-1)

F IGURE 2.1-6 2.1-48

9 BFS TABLE 2.1-1

(

REGIONAL INCORPORATED COMMUNITY STATISTICS Distance and 1970 g

County Otractlon Population inola Roge rs 3 miles NE 948 8 miles W5W 17 New Tulse Wagone r Fair Oaks Wa gone r 9 miles W W 23 Tiewsh*

Rogers to miles N 95*=

Catocss Rogers 12 miles WJ 973 Chouteau Mayes 13 elles ENE I.046 Coweta Wagone r 13 miles 55W 2.457 eroken Arrow Tulse ik miles WSW 11.787 C l aremre Rogers 14 miles Nw 9.084 Wagone r Wa gone r 15 miles SE 4.959 16 miles 5 230 Red Bird Wagone r Porter Wagone r 18 miles 5 624 P ryo r Mayes 19 miles NE 7.057 Owesso Tulse 10 miles Wd 3,491 Tullahassee Wa gone r 21 miles SSE 183 Maskell Muskogee 22 miles $$V 2,06)

Foyll Rogers 22 miles N I64 Bimby Tulsa 23 miles Sw 3,973 Locust Grove Mayes 23 miles ENE 1,090 Okay wagoner 23 miles SE 4I9 Tulsa Tulse 23 miles W 330,350 Collinsville Tulsa 24 miles Kd 3,009 Jecks Tulsa 15 miles wsw I.997 Taft Muskogee 25 mils 5 525 Oologeh Rogers 25 miles h%d 458

  1. eggs Cherokee 26 miles E 82 Salina Mayes 26 miles ENE I,024 Hulbert C he rckee 26 mile s ESE 505 Adair Meyes 23 miles NE 459
  • Tianah is en unincorporated area within 10 miles of the plant site.

It has been lectuoed in this I stir.g because of its proximity to the plant site.

(

    • Tlawah pcoulation is estimated f rom dwelling counts on the County Hig% ay Map insert.

2.1-9 I

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s BFS TABLE 2.1-3 LOCAL COMMUNITY POPULATION PROJECTION & DESSITY YEAR ESTIMATED POPULATION DENSI1T Inola (3 mi. NE) 1970 948 237 1974 1,176 345 1977 2,050 512 1980 2,900 725 1983 3,080 770 1990 3,700 925 2000 4,200 1,050 2010 4,450 1,112 2020 4,600 1,150 Tiawah (10 mi. N) 1970 95 127 1974 106 141 1977 116 155 1980 125 167 1983 135 180 1990 159 212 2000 213 284 2010 264 352

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