ML032100541
| ML032100541 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Watts Bar, Sequoyah  |
| Issue date: | 07/24/2003 |
| From: | Burzynski M Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML032100541 (64) | |
Text
Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402-2801 July 24, 2003 10 CFR 50, Appendix E Section V U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of Docket Nos. 50-259 50-390 Tennessee Valley Authority ) 50-260 50-391 50-296 50--327 50-328 TVA CENTRAL EMERGENCY CONTROL CENTER (CECC) - EMERGENCY PLAN IMPLEMENTING PROCEDURE (EPIP) REVISIONS In accordance with the requirements of 10 CFR Part 50, Appendix E,Section V, enclosed are copies of the Effective Page Listing and revisions to CECC EPIPs.
PROCEDURE EFFECTIVE DATE EPIP EPL 7/1/03 EPIP-8 Rev. 26 7/1/03 EPIP-13 Rev 10 7/1/03 If you have any questions, please contact Terry Knuettel at (423) 751-6673.
Sincerely, Mark J. Burzynski Manager Nuclear Licensing Enclosures cc: See page 2
- (--)lf5 Pnted r ecycd paper
U.S. Nuclear Regulatory Commission Page 2 July 24, 2003 cc (Enclosures):
U.S. Nuclear Regulatory Commission (Enclosures 2)
Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector [Enclosures provided Browns Ferry Nuclear Plant by site DCRM]
10833 Shaw Road Athens, Alabama 35611-6970 NRC Senior Resident Inspector [Enclosures provided Sequoyah Nuclear Plant by site DCRM]
2600 Igou Ferry Road Soddy Daisy, Tennessee 37379-3624 NRC Senior Resident Inspector [No enclosures, by request Watts Bar Nuclear Plant of site resident]
1260 Nuclear Plant Road Spring City, Tennessee 37381
DOCUMENT RELEASE AND FILING INSTRUCTIONS Page 1 of I Release No.
To: Management ServicesiRJNU1EDM Prepared By: Gall White Other Address: Extension: 751-2108 Date Submitted to Management Organization: AS&P Services/RIMWEDM:_ Address: LP 4D-C Date to Filed By:
Attached are: (select one) Release and Submitted for 0il QA Records/Documents 0 Distribution o Non-QA Records/Documents 0 Retention
_ _ ~~~REC NO. ACCPT REMOVE INSERT DOCUMENT NUMBER REV PAGES Y N DATE PAGES PAGES CECC-EPIP (6E-K0- 3)1-List of Effective Pages __ _ 07/01/03 1-9 3) -1 -9 CECC EPIP-8, cover sheet 26 1 _1 = 7-1-O3 cover sheet cover sheet CECC EPIP-8, rev. log 26 AZ rev. log rev. log CECC EPIP-8 26 32 _1 = All 1 -32 CECC EPIP-13, cover sheet 10 1 V 7- 1-3 cover sheet cover sheet CECC EPIP-13, rev. log 10 I rev. log rev. log CECC EPIP-13 10 15 All 1-15
___ ___ ___ Acceptance:
Date 0-7 WA= 407=(=7 ae1 fISP233(82-7 Contact Ext.
TVA 40074B (8-97) f08-971 Page of SPP-2.3-3 (08-29M
CECC-EPIP EPL Page 1 of 9 07/01/03 TENNESSEE VALLEY AUTHORITY CENTRAL EMERGENCY CONTROL CENTER EMERGENCY PLAN IMPLEMENTING PROCEDURES LIST OF EFFECTIVE PAGES This list of effective pages must be retained with the CECC-EPIPs.
Procedure No. Subdivision Page No. Rev. No.
List of Effective Pages 1 of 9 07/01/03 2 of 9 07/01/03 3 of 9 07/01/03 4 of 9 07/01/03 5 of 9 07/01/03 6 of 9 07/01/03 7 of 9 07/01/03 8 of 9 07/01/03 9 of 9 07/01/03 Table of Contents 1 of2 10/17/02 2 of 2 10/17/02 EPIP-1 Cover Sheet 37 Rev. Log 37 I of 33 37 2 of 33 37 3 of 33 37 4 of 33 37 5 of 33 37 Appendix A 6 of 33 37 Appendix B 7 of 33 37 8 of 33 37 Appendix C 9 of 33 37 Appendix D 10 of 33 37 Appendix E 11 of 33 37 Appendix F 12 of 33 37 Appendix G 13 of 33 37 14 of 33 37 Appendix H 15 of 33 37 Appendix I 16 of 33 37 17 of 33 37 18 of 33 37 19 of 33 37 Appendix J 20 of 33 37 21 of 33 37 22 of 33 37 23 of 33 37 Appendix K 24 of 33 37 25 of 33 37 26 of 33 37 Appendix L 27 of 33 37 Appendix M 28 of 33 37 Appendix N 29 of 33 37 30 of 33 37
CECC-EPIP EPL Page 2 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-1 (Continued) 31 of 33 37 32 of 33 37 33 of 33 37 EPIP-2 Cover Sheet 30 Rev. Log 30 1 of 5 30 2 of 5 30 3 of 5 30 Appendix A 4 of 5 30 Appendix B 5 of 5 30 EPIP-3 Cover Sheet 31 Rev. Log 31 I of 8 31 2 of 8 31 3 of 8 31 4 of 8 31 5 of 8 31 Appendix A 6 of 8 31 Appendix B 7 of 8 31 Appendix C 8 of 8 31 EPIP-4 Cover Sheet 32 Rev. Log 32 1 of 8 32 2 of 8 32 3 of 8 32 4 of 8 32 5 of 8 32 Appendix A 6 of 8 32 Appendix B 7 of 8 32 Appendix C 8 of 8 32 EPIP-5 Cover Sheet 34 Rev. Log 34 1 of 10 34 2 of 10 34 3 of 10 34 4 of 10 34 5 of 10 34 Appendix A 6 of 10 34 7 of 10 34 8 of 10 34 Appendix B 9 of 10 34 Appendix C 10 of 10 34 EPIP-6 Cover Sheet 24 Rev. Log 24 1 of 23 24 2 of 23 24 3 of 23 24 4 of 23 24
CECC-EPIP EPL Page 3 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-6 (Continued) Appendix A 5 of 23 24 6 of 23 24 7 of 23 24 Appendix B 8 of 23 24 Appendix C 9 of 23 24 Appendix D 10 of 23 24 11 of 23 24 Appendix E 12 of 23 24 13 of 23 24 14 of 23 24 Appendix F 15 of 23 24 Appendix G 16 of 23 24 17 of 23 24 Appendix H 18 of 23 24 19 of 23 24 Appendix I 20 of 23 24 21 of 23 24 Appendix J 22 of 23 24 Appendix K 23 of 23 24 EPIP-7 Cover Sheet 28 Rev. Log 28 1 of 15 28 2 of 15 28 3 of 15 28 4 of 15 28 5 of 15 28 6 of 15 28 7 of 15 28 8 of 15 28 Appendix A 9 of 15 28 Appendix B 10 of 15 28 Appendix C 11 of 15 28 Appendix D 12 of 15 28 13 of 15 28 Appendix E 14 of 15 28 Appendix F 15 of 15 28 EPIP-8 Cover Sheet 26 Rev. Log 26 1 of 32 26 2 of 32 26 3 of 32 26 4 of 32 26 5 of 32 26 6 of 32 26 Appendix A 7 of 32 26 Appendix B 8 of 32 26 Appendix C 9 of 32 26 10 of 32 26 Appendix D 11 of 32 26
CECC-EPIP EPL Page 4 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-8 (Continued) 12 of 32 26 Appendix E 13 of 32 26 14 of 32 26 Appendix F 15 of 32 26 16 of 32 26 17 of 32 26 18 of 32 26 19 of 32 26 20 of 32 26 21 of 32 26 22 of 32 26 Appendix G 23 of 32 26 Appendix H 24 of 32 26 Appendix I 25 of 32 26 26 of 32 26 27 of 32 26 28 of 32 26 29 of 32 26 Appendix J 30 of 32 26 31 of 32 26 32 of 32 26 EPIP-9 Cover Sheet 27 Rev. Log 27 I of 48 27 2 of 48 27 3 of 48 27 4 of 48 27 5 of 48 27 6 of 48 27 7 of 48 27 8 of 48 27 9 of 48 27 10 of 48 27 11 of 48 27 12 of 48 27 13 of 48 27 14 of 48 27 15 of 48 27 16 of 48 27 17 of 48 27 18 of 48 27 19 of 48 27 20 of 48 27 21 of 48 27 22 of 48 27 23 of 48 27 24 of 48 27 Appendix A 25 of 48 27 Appendix B 26 of 48 27 Appendix C 27 of 48 27
CECC-EPIP EPL Page 5 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-9 (Continued) 28 of 48 27 Appendix D 29 of 48 27 30 of 48 27 Appendix E 31 of 48 27 32 of 48 27 Appendix F 33 of 48 27 Appendix G 34 of 48 27 Appendix H 35 of 48 27 36 of 48 27 37 of 48 27 38 of 48 27 Appendix I 39 of 48 27 40 of 48 27 41 of 48 27 42 of 48 27 Appendix J 43 of 48 27 Appendix K 44 of 48 27 Appendix L 45 of 48 27 46 of 48 27 47 of 48 27 48 of 48 27 EPIP-11 Cover Sheet 12 Rev. Log 12 1 of 17 12 2 of 17 12 Appendix A 3 of 17 12 Appendix B 4 of 17 12 Appendix C 5 of 17 12 6 of 17 12 Appendix D 7 of 17 12 Appendix E 8 of 17 12 9 of 17 12 10 of 17 12 Appendix F 11 of 17 12 Appendix G 12 of 17 12 Appendix H 13 of 17 12 Appendix I 14 of 17 12 Appendix J 15 of 17 12 16 of 17 12 17 of 17 12 EPIP-12 Cover Sheet 18 Rev. Log 18 I of 19 18 2 of 19 18 Attachment 1 3 of 19 18 Attachment 2 4 of 19 18 Attachment 3 5 of 19 18 Attachment 3 6 of 19 18 Attachment 3 7 of 19 18 Attachment 3 8 of 19 18
CECC-EPIP EPL Page 6 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-12 (Continued) Attachment 3 9 of 19 18 Attachment 4 10 of 19 18 Attachment 5 11 of 19 18 Attachment 6 12 of 19 18 Attachment 7 13 of 19 18 Attachment 7 14 of 19 18 Attachment 8 15 of 19 18 Attachment 9 16 of 19 18 Attachment 9 17 of 19 18 Attachment 10 18 of 19 18 Attachment 10 19 of 19 18 EPIP-13 Cover Sheet 10 Rev. Log 10 1 of 15 10 2 of 15 10 3 of 15 10 4 of 15 10 5 of 15 10 6 of 15 10 Appendix A 7 of 15 10 8 of 15 10 9 of 15 10 Appendix B 10 of 15 10 Appendix C 11 of 15 10 12 of 15 10 13 of 15 10 14 of 15 10 Appendix D 15 of 15 10 EPIP-14 Cover Sheet 25 Rev. Log 25 I of 21 25 2 of 21 25 3 of 21 25 4 of 21 25 5 of 21 25 6 of 21 25 Appendix A 7 of 21 25 8 of 21 25 9 of 21 25 Appendix B 10 of 21 25 11 of 21 25 12 of 21 25 13 of 21 25 14 of 21 25 Appendix C 15 of 21 25 16 of 21 25 17 of 21 25 18 of 21 25 19 of 21 25 .
CECC-EPIP EPL Page 7 of 9 07/01103 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-14 (Continued) 20 of 21 25 Appendix D 21 of 21 25 EPIP-15 Cover Sheet 0 Rev. Log 0 I of 5 0 2 of 5 0 3 of 5 0 Appendix A 4 of 5 0 Appendix B 5 of 5 0 EPIP-17 Cover Sheet 17 Rev. Log 17 1 of 27 17 2 of 27 17 3 of 27 17 4 of 27 17 5 of 27 17 Appendix A 6 of 27 17 Appendix B 7 of 27 17 Appendix C 8 of 27 17 9 of 27 17 10 of 27 17 Appendix D 11 of 27 17 Appendix E 12 of 27 17 Appendix F 13 of 27 17 Appendix G 14 of 27 17 15 of 27 17 16 of 27 17 17 of 27 17 18 of 27 17 Appendix H 19 of 27 17 Appendix I 20 of 27 17 Appendix J 21 of 27 17 Appendix K 22 of 27 17 Appendix L 23 of 27 17 Appendix M 24 of 27 17 25 of 27 17 26 of 27 17 Appendix N 27 of 27 17 EPIP-18 Cover Sheet 9 Rev. Log 9 I of 6 9 2 of 6 9 3 of 6 9 4 of 6 9 Appendix A 5 of 6 9 Appendix B 6 of 6 9
CECC-EPIP EPL Page 8 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-19 Cover Sheet 12 Rev. Log 12 I of 11 12 2 of 11 12 3 of 11 12 4 of 11 12 5 of 11 12 Appendix A 6 of 11 12 7 of 11 12 Appendix B 8 of 11 12 9 of 11 12 10 of 11 12 11 of 11 12 EPIP-21 Cover Sheet 13 Rev. Log 13 I of 4 13 2 of 4 13 Appendix A 3 of 4 13 Appendix B 4 of 4 13 EPIP-22 Cover Sheet 19 Rev. Log 19 I of 7 19 2 of 7 19 Attachment A 3 of 7 19 Attachment B 4 of 7 19 5 of 7 19 6 of 7 19 7 of 7 19 EPIP-23 Cover Sheet 18 Rev. Log 18 1 of 24 18 2 of 24 18 3 of 24 18 4 of 24 18 Attachment A 5 of 24 18 Attachment B 6 of 24 18 7 of 24 18 8 of 24 18 9 of 24 18 10 of 24 18 11 of 24 18 Attachment C 12 of 24 18 Attachment D 13 of 24 18 Attachment E 14 of 24 18 15 of 24 18 Attachment F 16 of 24 18 17 of 24 18 of 24 18 18 L.
CECC-EPIP EPL Page 9 of 9 07/01/03 List of Effective Pages (Continued)
Procedure No. Subdivision Page No. Rev. No.
EPIP-23 (Continued) 19 of 24 18 20 of 24 18 Attachment G 21 of 24 18 22 of 24 18 Attachment H 23 of 24 18 Attachment I 24 of 24 18
CECC EPIP Coversheet Title CECC EPIP-8 Tennessee REV. 26 Valley Authority CENTRAL EMERGENCY DOSE ASSESSMENT STAFF AClVITIES CONTROL CENTER DURING NUCLEAR PLANT EMERGENCY PLAN RADIOLOGICAL EMERGENCIES IMPLEMENTING Effective Date:
PROCEDURES WRIlTEN BY: <5 REVIEWED BY: 7_ ____
Signature Signature D PLAN EFFECTIVENESS DETERMINATION: a 4La Signature Date 1
CONCURRENCES Concurrence Signature Date l ,EP Proa Planning and Implementation l Manager, ~rec rprSs l Marage'Radological and Chemistry Services 07fo1)°3 APPROVAL APPROVED BY: ___ VP Eng & Tech Svcs A E B:Signature Title Organization Date
CECC-EPIP-8 DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR PLANT RADIOLOGICAL EMERGENCIES REVISION LOG Rev. No. Date Revised Pages 0 3/22/88 All (Changed from IPD to EPIP) 1 11/18/88 2-7, Apps. A, B, C, & D 2 12/12/88 Appendix A 3 4/26/89 All 4 9/19/89 App. C 5 10/26/89 All 6 5/21/91 App. B, pgs. 1-4; Appendix C, pqs. 1-2; App. D, pg. 1 7 10/17/91 App. B, pgs. 2-4; App. C, pg. 1.
8 05/13/93 1-4; App. A; App. B, pg. 1, 3, & 4; and App. G; App. C deleted. All pages issued.
9 11/22/93 Pg. 4; App. B, pgs. 1&4; App. D changed to App. C; App. E changed to App. D; App. F changed to App. E; and App. G changed to App. F.
10 11/30/93 1, 3, 4; App. A, pg. 1; App. B, pgs. 1-2; App. C, pg. 1-5; App. D, pg. 1; App. E, pg. 1; App. F, pg. 1; App. G, pgs. 1-6.
11 06/24/94 App. B, pg. 1; App. D, pgs. 2-5; App. F; App. J added. All pages issued.
12 6/27/95 Pg. 1; App. A; App. B, p.3; App. C, p. 5; App. D, p. 2; App. G, pgs. 4 and 6 13 1/17/96 App. B, pg. 2, editorial changes, add table for BFN stack release; App. C, pgs. 1 & 3, Add new criteria for Type I and Type II releases; App. D, pgs. 2-5, add nonogram alignment checks 14 5/30/96 Pg. 3, App. A, App. B. App. C, App. D, App. F, App. G; annual review; ground level release tables and nomograms made generic to all three sites; all pages issued.
15 10/30/96 Pg. 3, App. B, and App. D; Add reference to App. I of CECC EPIP-7, remove deleted pages, make correction to Nomogram Alignment Check Table.
16 5/30/97 Editorial changes, update manual dose assessment methodology, update preliminary assessment table, revise river miles on tables in Appendix G, annual review. All pages issued.
17 8/8/97 Revise default river flow rate for BFN, revise responsibilities of Norris Lab, add water intake tables. All pages issued.
CECC-EPIP-8 DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR PLANT RADIOLOGICAL EMERGENCIES REVISION LOG (Continued)
Rev. No. Date Revised Pages 6/9/98 18 Annual review. Organization title changes. In Appendix D clarify RR Type I and Type II formulas. Remove Tennessee River miles from tables. All pages issued.
19 10/27/98 Correct reference to CECC EPIP-1 Appendix on Appendix J.
20 5/20/99 Annual review. Editorial and clarification changes, revise public water use tables. All pages issued.
21 9/8/00 Annual review. Editorial changes. All pages Issued.
22 3/30/01 Revised to incorporate the new source term methodology in the RED suite of codes revision Revised all pages to reflect human factor improvements In REP codes and manual Included changes due to code revision necessary for H-3 23 11/22/02 project.
24 3/31/03 Added sections to Appendix F to provide Instructions for manual method of calculating TEDE and thyroid CDE doses at Site Boundary (0.62 miles). All pages issued.
25 6116/03 Make corrections to Appendix F. All pages issued.
26 07/01/03 Revision consistency In accordance with CHPER 03-000257-000.
Revise location Indicator in step C.4 on page 18. Correct the location indicator, In step 2.b on page 19. Correct spelling of Circle on page 22.
Minor format alignment changes. All pages issued.
Table of Contents 1.0 PURPOSE ..... 2 2.0 SCOPE ........................ 2 3.0 STAFFING 3.1 Activation and Notification .2 3.2 Shift Change .2 3.3 Termination .2 4.0 DOSE ASSESSOR INTERFACES 4.1 Radiological Assessment Manager I Coordinator ......................................... 3 4.2 Meteorologist ..................................................... 3 4.3 Environs Assessor / Field Coordinator .................................................... 3 4.4 Core Damage Assessor ..................................................... 4 4.5 Technical Support Center ...................................................... 4 4.6 River Operations ...................................................... 4 4.7 State (Radiological) ..................................................... 4 5.0 PERFORMING DOSE ASSESSEMENTS 5.1 Data Verification ............. 4 5.2 Preliminary Assessments .................................................... 4 5.3 Criteria for Significant Change in Conditions ................................................ 5 5.4 FRED or RED Assessments - Collection of Data ......................................... 5 5.5 Preparing a Protective Action Recommendation (PAR) ................ ............... 5 5.6 Changes in Conditions for a PAR ...................................................... 5 5.7 BRED Assessment - Back Calculation of Release Rate from Field Data ..... 5 5.8 Comparison of Measured Field Data to Dose Projections ............................ 6 5.9 WATERDOSE Assessments .................. .................................. 6 5.10 Manual Methodologies for Dose Assessments ........................................ 6
6.0 REFERENCES
....................................... 6 7.0 ABBREVIATIONS ..... 6................................
Appendixes A. Dose Assessor Initial Reporting Checklist ................................................ 7 B. Shift Change and Termination Checklist ............................................... 8 C. FRED RED Data Inputs ................................................ 9 D. FRED / RED Assessment of Airborne Releases ........................................ 11 E. BRED Assessment of Airborne Release Field Data ................................... 13 F. Manual Methodology for Assessing Airborne Releases ............................. 15 G. Comparison of Measured Field Data to Dose Projection Models ............... 23 H WATER DOSE Data Inputs ............................................... 24 I. Manual Methodology for Assessing Uquid Releases to the River .............. 25 J. Public and Industrial Surface Water Supplies ............................................ 30
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 2 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR PLANT RADIOLOGICAL EMERGENCIES 1.0 PURPOSE To guide Dose Assessment in obtaining necessary information, calculating dose rates and doses, and communicating assessment results used in responding to radiological emergencies at nuclear power plants.
2.0 SCOPE This procedure applies to activities of Dose Assessment in actual and hypothetical radiological emergency situations. While the activities of the Dose Assessment staff are expected to follow this procedure, it is expected that circumstances may arise during an event which will void portions of this procedure. Therefore, this procedure is a guide for the operation of the Dose Assessment staff under the ideal conditions.
3.0 STAFFING 3.1 Activation and Notification The Initial notification of an event comes from the Operations Duty Specialist via the Emergency Paging System (EPS) or manual callout. Additional Dose Assessor support is contacted in accordance with Appendix A. The Dose Assessor is a position required for the CECC to make Protective Action Recommendations and to meet minimum staffing levels.
Upon reporting to the CECC, perform initial activities in accordance with the checklist provided as Appendix A.
3.2 Shift Change Shift change notification and transition and transfer of responsibilities should be conducted in accordance with the Dose Assessment Shift Change and Termination Checklist (Appendix B).
3.3 Termination Termination of an event should include the following actions and follow the Dose Assessment Shift Change and Event Termination Checklist (Appendix B).
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 3 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES 4.0 DOSE ASSESSOR INTERFACES 4.1 Radiological Assessment Manager I Coordinator (RAMIRAC)
The Dose Assessor should interface directly with the Radiological Assessment Coordinator (RAC). In the absence of the RAC, communication Is provided directly to the RAM. Requests for any special-case assessments should come to Dose Assessment through the RAC/RAM or be cleared by the RAC/RAM prior to their performance.
Dose Assessment is responsible for performing the offsite dose assessment activities of the CECC in order to determine Protective Action Recommendations using the appropriate appendix in CECC EPIP-1. Dose assessment results are also evaluated against criteria for declaration of Emergency Classification levels, and evaluations are communicated to the RAM/RAC.
Dose Assessment should provide results of all dose assessments and plume plots from FRED to the RAC/RAM, who will approve them and distribute them to CECC staffs. Initial dose assessments (those made at the start of an event or when the conditions have changed significantly as defined in this procedure) will receive the approval of the RAC/RAM and then are transmitted to the TSC and the State. Under most other conditions, the results are directly transmitted to the TSC and the State on the State Update form via computer spooling. However, if the computer spooling is unavailable, then the Dose Assessor shall prepare a State Update form manually as defined in EPIP-1. CECC Clerical staff have Instructions for distribution.
Dose Assessment should provide to the RAC/RAM copies of plume plots from RED for ongoing releases or plots of estimated centerline location (if there is not a known release but potential exists for one to occur). This Information should be transmitted by the RAC/RAM to the CECC, TSC, and State. Dose assessment will also support post event recovery efforts.
4.2 Meteorologist (MET)
The CECC Meteorologist Is responsible for providing to Dose Assessment the real time and forecast meteorological data and associated advice on atmospheric dispersion and transport.
If a meteorologist is not initially available for response to the CECC, support can be obtained from Muscle Shoals. Telephone and pager numbers for the Muscle Shoals response personnel are available Inthe REND.
Meteorological data is provided to the CECC by computer inputs and by the CECC meteorologist. In the event of a monitored airborne release, the 15-minute meteorological data is automatically accessed by the RED and FRED codes. This data should be verified against the distributed meteorological data or by the meteorologist. The meteorologist Isalso available to convert flow rates to exit velocities for use Inthe codes. The meteorologists will also provide forecast information for use in the FRED code.
4.3 Environs AssessorlField Coordinator Dose Assessors provide plume plots to the CECC Environs Assessor and to the Field Coordinator at the Radiological Monitoring Control Center (RMCC) via the RAC. These plume plots are used to assist with decisions on field team deployments. Real time plume plots from the RED code are to be distributed to the EA/FC and the State for that purpose.
Field data is also shared to assist with comparison of dose projections with field measurements. This comparison can assist with evaluations if field teams are at maximum centerline locations, or If reported plant release rates coincide with actual field measurements.
I In the event of an unmonitored release from a site, field team data can be used in the BRED code to assist with determination of a release rate.
4.4 Core Damage Assessors The CECC Core Damage group (in Plant Assessment) is responsible for supplying Dose Assessment with projections of potential, anticipated, and/or worst-case release rates and pathways.
4.5 Technical Support Center (TSC)
The TSC is a source of information for radioactivity release rates, pathways, flow rates, and information on plant status and prognosis. The primary point of contact is TSC Chemistry.
Release information is also available via the Integrated Computer System (ISC) using the CECC computers.
4.6 River Operations River Operations may assist in providing Dose Assessment information on water dispersion characteristics for releases to the river. This information may be used in running the WATERDOSE code, or for use of the manual methodology if the dose code is unavailable.
4.7 State (Radiological)
Dose Assessment shall ensure that communication with the State Dose Assessment Team is established and maintained. The State should be given hourly updates, as a minimum. These updates should include discussions of all technical information relative to dose assessments being made (incoming release rates, assumptions used, problems with information flow). The State should also be contacted if the conditions have changed significantly as defined in this procedure. DO NOT discuss protective action recommendations with the State.
5.0 PERFORMING DOSE ASSESSMENTS 5.1 Data Verification All dose assessment results (computer generated or hand calculated) involving data input will be verified by a second party verifier. The verifier may be a Dose Assessor or the RAM/RAC. The verifier will verify the accuracy and appropriateness of data input and reasonableness of the results. Both preparer and verifier will initial and date the results page of the assessment (e.g.,
State Update Form for FRED assessments).
5.2 Preliminary Assessments Dose Assessors should provide results of all preliminary assessments to the RAC/RAM.
Preliminary Assessments are provided as part of a FRED run. Preliminary assessments will be performed at the start of an event or when the conditions have changed significantly as defined in this procedure.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 6 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES 5.3 Criteria for a Significant Change in Conditions Criteria for a significant change which will require a new dose assessment run are:
- the release type I path has changed,
- the release rates have changed by a factor of 1o
- the stability class has changed by 2 classes,
- or the wind speed has changed by a factor of 2.
5.4 FRED or RED Assessments - Collection of Data Gather information as provided on Appendix C. Sources of information may Include the Technical Support Center (Chemistry), ICS, CECC Meteorologist or CECC Core Damage Assessors. Refer to Appendix C for instructions on running the dose codes.
5.5 Preparing a Protective Action Recommendation (PAR)
TVA must satisfy regulatory requirements to provide State Authorities a PAR within 15 minutes of the declaration of a General Emergency. Therefore, Dose Assessors should anticipate and initiate development of a PAR to allow ample time for review, approval and transmittal to State Authorities.
A Protective Action Recommendation for airborne releases is determined based upon results of a FRED run. If the FRED program Is unavailable, then the manual methodology should be utilized as provided in this procedure. A PAR form contained in CECC EPIP-1 should be completed, with attention to identification of affected sectors as page 2 of that document..
Dose Assessment should provide technical guidance to the RACIRAM Inthe preparation of protective action recommendations based on dose assessments. The RAC is responsible for written preparation of recommendations to the RAM.
5.6 Changes in Conditions for a PAR Changes to a PAR must be communicated to the State by the CECC Director within 15 minutes of determination. Criteria for a changes which will require evaluation a new PAR are:
- the release type / path has changed,
- the release rates have changed by a factor of 10
- the stability class has changed by 2 classes,
- or the wind speed has changed by a factor of 2.
- a wind direction change resulting Ina change of an affected sector 5.7 BRED Assessment - Back Calculation of Release Rate from Measured Field Data Measured field data (consisting of dose rates in mrem/hr and 1-131/1+3 concentrations) are assessed in several ways. If there Is a monitored release ongoing, the field data are compared to the results of the most applicable data produced by the RED or FRED computer models.
However, in cases where the release is unknown or questionable, the field data are then input Into the BRED computer model to determine the applicable release rates. These calculated release rates are then Input into the RED/FRED codes, as applicable, which can be used to perform dose assessments and any applicable Protective Action Recommendation (PAR).
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 6 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES 5.8 Comparison of Measured Field Data to Dose Projections Field data is compared with dose projections to assist with evaluations if field teams are at maximum centerline locations, or if reported plant release rates coincide with actual field measurements. Appendix G is provided as a reference to perform comparisons.
5.9 WATERDOSE Assessments Liquid releases to the River are assessed using the WATER DOSE code as provided on Appendix H. If the WATER DOSE code is unavailable, a manual methodology is provided as Appendix I.
5.10 Manual Methodologies for Dose Assessments In the event that the FRED, RED or WATERDOSE computer codes are unavailable, instructions are provided in the Appendixes of this procedure for manual calculation methods. In consideration that the computer programs also normally spool data outputs directly to the State, the Dose Assessor will need to ensure that the applicable pages of the State Update Form, contained In CECC EPIP-1, are also manually completed and transmitted accordingly.
6.0 REFERENCES
FRED User's Manual RED/FRED/BRED Documentation FRED User's Manual WATERDOSE User's Manual BRED User's Manual Model Comparison REP CODE Revision 2, Specifications and Documentation, August 2002, L61 020814 800 7.0 ABBREVIATIONS AND DEFINITIONS CECC - Central Emergency Control Center CTM - Containment building SGTR (above) - Steam Generator Tube Rupture above the steam generator water level SGTR (below) - Steam Generator Tube Rupture below the steam generator water level MSLB - Main Steam Line Break TSC - Technical Support Center EPS - Emergency Paging System RED - Radiological Emergency Dose Code RO - River Operations FRED - Forecast Radiological Emergency Dose Code BRED - Back-calculation Radiological Emergency Dose Code TRM - Tennessee River Mile ICS - Integrated Computer System WGDT - Waste Gas Decay Tank (as in rupture event)
RAM/RAC -Radiological Assessment Manager or Radiological Assessment Coordinator
APPENDIX A Dose Assessor Initial Reporting Checklist (steps do not need to be performed Insequential order)
- 2. START logkeeping of key activities and notifications in the position logbook.
- 3. ENSURE that the following support staffs are notified and/or staffed. Refer to the REND call out list for contact information.
- Second Dose Assessor, if needed.
- Muscle Shoals Meteorologist (if serving as CECC pager duty person).
- 4. CONFIRM position notebook procedures match revision levels in controlled copies.
- 5. ESTABLISH contact with the TSC Chemistry (programmed on phone and in REND section B). Ascertain if a release has been, or is occurring. IF YES, INITIATE a dose assessment as noted below.
- 6. Perform preliminary assessments and dose projections.
- 7. ESTABLISH initial contact with the State Radiological Dose Assessment staff (programmed on phone and in REND section B).
- 8. OBTAIN a briefing from the RAC/RAM and INFORM the RAC/RAM when the activities above are completed. Report/request if a radiological release has been, or is occurring.
NOTES: COMPARE dose assessment results against the levels for the declared REP class and advise the RAC/RAM to advise the TSC if an upgrade is indicated.
For Preliminary Assessments and Dose Projections use the FRED Code (Appendix C and D).
For Plume Plots to track actual releases In current time, use the RED Code (Appendix C and D).
When the plant release rate is unmonitored or questionable, use the BRED code to arrive at a plant release rate based upon Field Team data. (Appendix E).
For releases to the River, use the WATERDOSE Code (Appendix H)
Wcomputer problems are encountered. Immediately contact Comouter Support.
If the FRED computer code Is Inoperative, use the MANUAL METHODOLOGY to assess airborne radioactivity releases (Appendix F).
If the WATERDOSE computer code is Inoperative, use the MANUAL METHODOLOGY to assess liquid releases to the river (Appendix I and J).
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 8 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES APPENDIX B Dose Assessor Shift Chance and Termination Checklist
- 1. The following should be discussed between staff for Shift Turnover.
- Current release data and projections.
- Current met data and projections.
- Current plant status and projections.
- Current environs data and projections.
- Pertinent historical data/plant conditions
- Status of any Protective Action Recommendations made and the rationale for these
- Status of any (incoming or outgoing) unfulfilled requests for information.
- Dose methodologies being used.
- Identification of problems in response capability.
- Time for next periodic update to the State
- Time for next periodic update of the RED plume plot
- Identify Individuals external to CECC who were activated or placed on standby
- 2. Transfer of Shft Change Responsibility
- Obtain approval from the RAC for the transfer of responsibility
- The on duty Dose Assessment Staff should remain available or at least respond in case transfer problems are identified
- 3. Termination
- Log off CECC computer system/turn off plotters.
- Notify all on-call staff of event termination, such as:
- Meteorologist (if staffed in Muscle Shoals)
- Additional Dose Assessment staff on standby
- River Operations
- Collect and turn in all records to the EP staff
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 9 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 E MERGENiCIES:_
Appendix C Page 1 of 2 FRED RED FRED / RED Data Inputs Data Inputs NOTE: The source for this information may be the site Technical Support Center or from ICS.
- 2. Meteorological data will be: Ea ACTUAL or El EXERCISE (confirm with drill controller).
- 3. Release start time: El Eastern E] Central
- 4. Elapsed Time from reactor shutdown to start of release: _ (hours) (enter 0 if Rx under power)
- 5. Release Vent Type (this is used by the code to calculate effective plume height):
SQN/M/BN BFN a Shield Bldg El Stack
[I Near ground E Radwaste Zone (of Rx Bldg)
ED Refueling Bldg zone (of Rx Bldg)
E] Reactor Bldg zone (of Rx Bldg)
El Turbine Bldg zone (of Rx Bldg)
El Near ground
- 6. Effluent flow rate (exit speed) (if measured and available): cfm.
NOTE: Consult with the meteorologist as to whether the default Exit Velocity based on this flow rate should be over-ridden. Code defaults can be used for conservatism or if flow data is unavailable.
- 7. Release Type:
EJRCS []Core Damage []User Specified
[lGap (default) [:Fuel Melt (for noble gas and Tritium only)
NOTE: Initially, a GAP Release Type should be used unless otherwise specified by the Core Damage Assessment team. Alternately, particulate-to-1131field team air concentration data can be used as follows:
Field Team Data Particulate microCilcc = Ratio Field Team Data odine"" microCi/cc Ratio = Release Type: Core Damage Fuel Melt 2-0.18 2t2.0 2 3.5
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 10 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES Appendix C Page 2 of 2 FRED l RED FRED / RED Data Inputs Data Inputs
- 8. Release Path:
SQN/WBN
[] Filtered via containment (CTM)
El Unfiltered via containment (CTM)
El SGTR with rupture located BELOW water level E Steam Generator Tube Rupture with rupture located ABOVE water level El Turbine Bldg El Reactor Bldg El Auxiliary Bldg.
BFN El Stack (filtered) El Turbine Bldg, Reactor Bldg El Stack (unfiltered) E Main Steam Line Break (MSLB)
- 9. Release rates:
Basis for rates: E Monitor reading E Plant personnel El BRED estimate
..Ci/s Noble Gas Ci/s 1-131 (pre-treatment value only, if available)
Ci/s Total Particulate (pre-treatment value only, if available)
Ci/s H-3 (if applicable see note below)
NOTE:
- For a TPBAR handling accident, the H-3 release can be estimated as:
H-3 Release t.i/,.r.HER of-Aft.-w11.
A rfm
, fil eA 73'D3 R<LV
-1rrf L. L.
Rate 60 min/sec
{ H-3 release (Ci/s) = #iLCicc H-3 x building exhaust flow rate (cfm) x 28320 cc/cf x 1/60 min/s )
- For a WGDT Rupture accident, the default H-3 release is 2500 Ci over one hr or 6.94E+05 RC/s for 1 hr
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 11 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26
_ EMERGENCIESl Appendix D Page 1 of 2 FRED RED FRED or RED Assessment of Airborne Releases Code Runs
- 1. DOUBLE CLICK on the "CECC VAX' icon if the VAX User Window is not displayed on computer screen. Depress [RETURN] until prompted for the user name.
- 3. FOLLOW computer prompts to begin or exit program.
INOTE: TYPE CTRL Z any time to exit or re-start program.
When executing the RED code you will be asked whether this is a unew run." ANSWER U( and ENTER NEW RUN," unless you desire to modify or append to a current run.
- 4. INPUT data as collected on Appendix C.
For a user-specified release (for noble gas and/or tritium releases only), ENTER the nuclide number below (as applicable) and the associated nuclide-specific release rates.
Nuclide # Nuclide Nuclide # Nuclide I H-3 28 XE-131M 6 KR-85 29 XE-1 33 7 KR-85M 30 XE-133M 8 KR-87 31 XE-135 9 KR-88 32 XE-138
- 5. CONFIRM whether the release rate data is correct, (YIN). Edit as necessary.
6 CONFIRM whether the calculated release rate data is correct, (Y/N). Edit as necessary.
- 7. RUN the code for the expected event duration:
a For FRED Preliminary Assessments use 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; For FRED Dose Projections use a 4-hour duration unless known otherwise.
For RED assessments run once per 15-min during ongoing releases.
Appendix D Page 2 of 2 FRED l -
RED FRED or RED Assessment of Airborne Releases Code Runs
- 8. OBTAIN code outputs by as follows:
- a. ANSWER "Y" to the prompt to Print dose charts or plume plots."
- b. SELECT State Update Form (SUF) and plume plot as minimum outputs
- c. SELECT scale to be used:
[1] for 10 mile, [2] for 50 mile, [3] to exit code or go to next time segment)
- d. For plume plot, CLICK print button at bottom of screen to perform a screen print of plot.
Be sure that the pop-up dialog box has the Graphic Image set to "Swap Black/White."
- e. For Preliminary Assessments, OBTAIN the Protective Action Guide (PAG) release rates from the FRED output and the actual/projected release rates from the State Update Form.
- f. The Preparer and Verifier shall INITIAL and DATE the results.
- 9. COMPARE the declared REP class with that indicated in the FRED output. Notify the RAC/RAM (to advise the TSC) of the need for REP class changes based on radiological conditions.
- 10. GIVE PAG and actual/projected release rates to the Board Writer.
- 11. GIVE the FRED results (SUF, PAG Release Rates, plume plot, and REP class information) to the RAC for distribution. (The SUF may be sent directly through the computer to the State and the TSC.)
- 12. At the request of the RAC/RAM, PREPARE a PAR using the CECC Protective Action Logic Diagram and the PAR form found in EPIP-7 and give to the RAM with the results of the FRED run.
- 13. REQUEST that the RAC distribute the SUF, and any plume plots to all standard distribution locations, via CECC Clerical instructions.
- 14. Preferably once every 15-min (at least once per hour) during an actual release,
- a. ENTER the release data into the RED code for use in tracking the plume
- b. COMPARE the estimated impacts to measured field data.
- c. GIVE the results (plume plot only) to the RAC for distribution to the CECC, the State, and the TSC.
- 15. TYPE CTRL Z any time to exit or re-start program.
Appendix E Page of 2 BRED BRED Evaluation of Airbome Field Data Code Run
- 1. Log on to BRED. DOUBLE CLICK on the "CECC VAX" icon. PRESS return until prompted for usemame. ENTER usemame (BRED) and password (CECC).
- 2. OBTAIN the following field data from Environs Assessment.
NOTE: As a minimum, only need one of the following measurements:
Dose Rate OR odine-131 OR Tritium (H-3)
I Distance (iles I Directio (secto) l-- TmeTTaken-D OR te-mrem/h (1 meter w/c)
I
............ ine
- 3. Elapsed Time from reactor shutdown 1 IC Y& Conc.....t.... Ci..
I to time of field measurement: (hours) (enter 0 if Rx under power)
- 4. DETERMINE the Release Path:
SQN/WBN o Filtered via containment (CTM)
O Unfiltered via containment (CTM) o SGTR with rupture located BELOW water level o Steam Generator Tube Rupture with rupture located ABOVE water level o Turbine Bldg ol Reactor Bldg
[ Auxiliary Bldg.
BFN o Stack (filtered) o Turbine Bldg, Reactor Bldg o Stack (unfiltered) o Main Steam Line Break (MSLB)
- 5. DETERMINE the Release Type: ORCS OCore Damage E]Gap (default) OFuel Melt NOTE: Initially, a GAP Release Type should be used unless otherwise specified by the Core Damage Assessment team. Alternately, particulate-to- 113 ' field team air concentration data can be used as follows:
Field Team Data Particulate microCi/co = Ratio Field Team Data lodine"' microCi/cc Gao Cor Damage Fuel Melt Ratio = Release Type: Z 0.18 > 2.0 t 3.5
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 14 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES l
Appendix E Page 2 of 2 BRED Code Run BRED Evaluation of Airborne Field Data
- 6. RUN the BRED computer model and follow prompts using the information from sections 2-5.
TYPE CTRL Z any time to exit or re-start program.
- 7. RECORD the printed Release Rate output (as applicable) and INPUT into the FRED code.
Noble Gas Cis) l Iodine 131 (pre-treatment) LCiIs Tritium 3 H) IIIs
- 8. COMPARE the new FRED run dose rate output to the previous RED/FRED computer model by CALCULATING a data ratio as follows:
FRED or RED Centerline Dose Rate dividedby the FIELD DATA Centerline Dose Rate divided by RATIO Unlikely - Field Team Plant but release possible not on unidentified?
NOBLE GAS RATIO Not unusual centerline?
I I I I 11I I 0.001 0.01 0.1 0.2 0.5 1 10 I I 50 100 I
1000 Base actions on Base Attempt to verify projections by verified field actions traversing the plume. Ask for measurements on reevaluation and level of projections confidence on plant release data. Must be very confident that field data is correct to override projections based on measured release data.
- 9. PROVIDE feedback to the Environs Assessor and RAC/RAM. UPDATE dose projections as necessary and give the results to the RAC for use in preparing Protective Action Recommendations (PAR), or prepare a PAR in accordance with CECC EPIP-1.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR PLANT RADIOLOGICAL EMERGENCIES CECC EPIP-8 Page 15 of 32 Revision 26 I Appendix F Page of 8 MANUAL AIR Manual Method for Assessinq Airborne Releases SB TEDE PAG A. Calculating TEDE PAG Release Rate at SITE BOUNDARY 0.62 miles) I
Stability Class: (circle) A B C D E F G Release Type: CIRCS [Core Damage []User Specified E[Gap (default) E]Fuel Melt (for noble gas and Tritium only)
Release Path: SQNIBN BFN O Filtered via CTM O filtered via stack O Unfiltered via CTM El unfiltered via stack EO SGTR with rupture BELOW water E] Turbine, Reactor Bldg O SGTR with rupture ABOVE water El Main Steam Line Break El Turbine, Reactor, Auxiliary Building
NOTE: USE ground level for all cases except for BFN stack.
A B I C D E F G Ground 1.7E+09 I 4.8E+08 2.2E+08 1.1 E+08 7.4E+07 4.9E+07 2.9E+071 Stack 1.8E+09 9.1E+08 9.1E+08 8.3E+08 8.3E+08 8.OE+08 9.1E+08
- 3. CIRCLE the appropriate TEDE Ratio below, based on release type/path:
TEDE Ratio (for 0.62 ml) BWR PWR Core Fuel User RCS Gap Damaae Melt Spec Stack (unfiltered 2.0 N/A 1.8 1.3 2.0 1.0 Stack (filtered) 1.9 N/A 1.0 1.0 1.0 1.0 CTM (unfiltered) or SGTR NIA 7.4 9.0 - 5.3 11 1.0 (below)
CTM (filtered) N/A 3.7 1.0 0.9 1.0 1.0 SGTR (above water) N/A 95 221 111 263 1.0 MSLB (BFN) 7.4 N/A 84 44 100 1.0 Turbine, Reactor or Aux Bldg 4.4 17 32 16 37 1.0
X /
TEDE PAG FACTOR wind speed TEDE Ratio TEDE NGPAG Release Rate (gCi/m, item 2) (mis, item 1) (item 3) SB 0.62 ml (Ci/s)
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 16 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES Appendix F Page 2 of 8 MANUAL AIR l-Manual Method for Assessing Airborne Releases SB TEDE PAG
- 5. OBTAIN the actual/projected Noble Gas Release Rate pci/s.
THEN radiological conditions indicate a General Emergency.
For Tritium Accidents (e.g., TPBAR handling or WGDT rupture),
l A l B C D E l F G I 4.OE+09 8.7E+08 2.9E+08 1.OE+08 5.9E+07 3.1 E+07 11 .4E+07
Tritium PAG FACTOR wind speed TEDE Tritium PAG Release Rate (pCi/m, item 7) (m/s item 1) SB 0.62 Ml (lCi/s)
- 9. OBTAIN the actual/projected Tritium Release Rate (see below) p.Cis.
NOTE:
- For a TPBAR handling accident, the H-3 release can be estimated as:
H-3 Release = uCifcc H-3 x cfm x 28320 cclcf Rate 60 min/sec
{ H-3 release (Ci/s) = #RlCi/cc H-3 x building exhaust flow rate (cfm) x 28320 cc/cf x 1/60 min/s }
- For a WGDT Rupture accident, the default H-3 release is 2500 Ci over one hr or 6.94E+05 gCi/s for 1 hr
THEN radiological conditions indicate a General Emergency.
- 11. IF tritium accident also involves noble gases, THEN perform the following calculation:
NG Release Rate + Tritium Release Rate TEDE NG PAG Release Rate TEDE Tritium PAG Release Rate (item 5) (item 9) =
+
(item 4) (item 8) Ratio
- 12. IF the value in item 11 2 1.0. THEN radiological conditions indicate a General Emergency.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 17 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26
_EMERGENCIES__ _ _
Appendix F Page 3 of 8 MANUAL AIR Manual Method for Assessing Airborne Releases I S THYROID B. Calculating THYROID CDE PAG Release Rate at SITE BOUNDARY (0.62 MILES)
- 1. USE the data from section A.1
NOTE: USE ground level for all cases except for BFN stack.
A B C D E F G Ground 4.1E+05 8.6E+04 3.OE+04 l 1.OE+04 6.OE+03 3.1E+03 1.4E+03 Stack 4.1 E+05 2.3E+05 3.7E+05 6.5E+05 1.4E+06 2.1 E+07 4.9E+11
x _
CDE PAG FACTOR wind speed CDE PAG Release Rate (RCi/m, item 2) (mS SIB 0.62 mi (Ci/s)
- 4. a. If known, RECORD the actual/projected 1-131 release rate pCi/s and go to Step 5.
If unknown, CIRCLE the 1-131 to NG ratio below, based on release type and path and continue with step 4 b.
.. :,sS,..3 _o as _ K3Ke s~~~~ il-tere' hi~O ___C 3.O . . orilrnaePuls 1..EAS A EO5 4.b. CALCULATE actual/projected iodine-131 release rate as follows:
X_
ActuaLUProjected NG release rate 1-131 to N ratio Actuaproj.1-131 release rate (item A.5) (item 4a) (9Ci/s)
THEN radiological conditions indicate a General Emergency.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 18 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES Appendix F Page 4 of 8 MANUAL AIR Manual Method for Assessing Airborne Releases I SB TEDE RATE IC. Calculating TEDE Dose Rate at SITE BOUNDARY 0.62 miles)
Stability Class: (circle) A B C D E F G Release Type: ORCS OCore Damage OUser Specified (for noble gas and Tritum only)
OGap (default) OFuel Melt Release Path: SQN/WBN BFN o Filtered via CTM o filtered via stack O Unfiltered via CTM o unfiltered via stack o SGTR with rupture BELOW water o Turbine, Reactor Bldg o SGTR with rupture ABOVE water ol Main Steam Line Break O Turbine, Reactor, Auxiliary Building Noble Gas Release Rate: - (Cis)
- 2. CIRCLE the TEDE FACTOR (rem/h per pCi/m) below, based on the stability class and release level.
NOTE: USE ground level for all cases except for BFN stack.
A l B C lD I El F G Ground 6.0E-10 2.1E-09 4.6E-09 9.5E-09 I 1.4E-08 2.1E-08 3.5E08 Stack 5.5E-10 1.IE-09 1.IE-09 1.2E-09 1.2E-09 1.3E-09 1.AE-09
- 3. CIRCLE the appropriate TEDE Ratio below, based on release type/path:
TEDE Ratio at 0.62 mil BWR PWR Core Fuel User RCS RCS GAP - Damage Melt Spec Stack (unfiltered) 2.0 N/A 1.8 1.3 2.0 1.0 Stack (filtered) 1.9 N/A 1.0 1.0 1.0 1.0 CTM (unfiltered) or SGTR (below) N/A 7.4 9 5.3 11 1.0 CTM (filtered) N/A 3.7 1.0 0.9 1.0 1.0 SGTR (above water) N/A 95 221 111 263 1.0 MSLB (BFN) 7.4 N/A 84 44 100 1.0 TB, RxB, AB 4.4 17 32 16 37 1.0
- 4. CALCULATE the TEDE Dose as follows:
X X I NG release rate (pCi/s) TEDE FACTOR TEDE Ratio iwind sp. (m/s) TEDE (rem/h)
(item A-5) (item 2) (item 3) (tem 1) 0.62 mile For Tritium Accidents (e.g., TPBAR handling or WGDT rupture),
I A B C D E I 2.5E-10 1.2E-09 3.5E-09 1.OE-08 1.7E-08 3.3E-08 7.OE-08
- Revision
Appendix F Page 5 of 8 Manual Method for Assessing Airborne ReleasesI MANUAL AIR mi TEDE RATE
- 6. CALCULATE the Tritium TEDE as follows: 5 ml THY CDE RATE X l Tritium Release Rate* Tritium TEDE FACTOR wind speed Tritium TEDE (lcis) (item 5) (ms) (rem/h)
(item 1)
- NOTE:
- For a TPBAR handling accident, the H-3 release can be estimated as:
H-3 Release =_ Ci/cc H-3 x cfm x 28320 cc/cf Rate 60 min/sec
( H-3 release (Ci/s) = #gCi/cc H-3 x building exhaust flow rate (cfm) x 28320 cc/cf x 1/60 minis
- For a WGDT Rupture accident, the default H-3 release is 2500 Ci over one hr or 6.94E+05 gCis for 1 hr
+
TEDE (rem/h) Tntium TEDE (remn/h) Total TEDE (rernmh) 0.62 mile D. Calculating SB THYROID CDE Dose Rate
- 1. CIRCLE the Thyroid CDE FACTOR (rem/h per gCi/m), based on the stability class and release level.
NOTE: USE ground level for all cases except for BFN stack.
A B C EE G Ground 1.2E-05 5.8E-05 1.7E-04 4.8E-04 8.3E-04 1.6E-03 3.5E-03 Stack 1.2E-05 2.2E-05 1.4E-05 7.7E-06 3.5E-06 2.3E-07 1.OE-1 1 2 a. If known, RECORD the 1-131 release rate _Cis and go to Step 3.
If unknown, CIRCLE the 1-131 to NG ratio below, based on release type and path and continue Ufih ci0n Vh 2b. CALCULATE actual/projected odine-131 release rate as follows:
X -_=.
NG release rate 1-131 to NG ratio 1-131I release rate (Ci/s)
- (item A.5) (item 2a)
Revision
- 3. CALCULATE Thyroid CDE Dose Rate as follows:
_______ X I 1-131 release rate (j+/-Ci/s) Thyroid CDE FACTOR wind speed (item 2) (item 1) (m/s)
IE. Calculating 6 mile TEDE USE the data from section A.1, THEN
- 1. OBTAIN an estimate of the release duration (t) hours. Use 4 (four) hours unless known otherwise.
- 2. CIRCLE the TEDE FACTOR (rem/h per per tCilm) below, based on the stability class and release level.
NOTE: USE ground level for all cases except for BFN stack.
A I B I C I D I E I F I G Ground 9.5E-11 I 1.5E-10 2.8E-10 I9.5E-10 I1.8E-09 3.5E-09 6.5E-09 Stack 9.OE-11 1.5E-10 2.6E-10 7.5E-10 1.1E-09 1.3E-09 1.1E-09
- 3. CIRCLE the appropriate TEDE Ratio below, based on release type/path:
TEDE Ratio at 5 min BWR PWR Core Fuel User RCS RCS GAP Damage Melt Spec Stack (unfiltered) 2.1 N/A 2.8 1.9 3.1 1.0 Stack (filtered) 2.1 N/A 1.0 0.9 1.0 1.0 CTM (unfiltered) or SGTR (below) N/A 3.5 4.9 2.9 5.8 1.0 CTM (filtered) N/A 1.8 1.0 1.0 1.0 1.0 SGTR (above water) N/A 43 100 51 116 1.0 MSLB (BFN) 4.5 N/A 40 21 47 1.0 TB, RxB, AB 3.1 7.4 15 7.9 17 1.0
- 4. CALCULATE the TEDE Dose as follows:
X X X NG release rate (Ci/s) TEDE FACTOR TEDE Ratio Duration (hrs) wind sp. (m/s)
(item A.5) (item 2) (item 3) (item 1) (item 1)
For Tritium Accidents (e.g., TPBAR handling or WGDT rupture),
I l B I _ D lF G l l 4.OE-11 5.OE-11 1.1E-10 4.4E-10 9.5E-10 2.4E-09 5.5E-09
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 21 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES___
Appendix F Page 7 of 8 MANUAL AIR Manual Method for Assessing Airbome Releases 5 m TEDE DOSE
X X I =,
Tritium Release Rate Tritium TEDE FACTOR Duration wind speed Tritium TEDE (OCis) (item 5) (hrs) (Ms) (rem)
(item A.1) (item A.1)
NOTE:
- For a TPBAR handling accident, the H-3 release can be estimated as:
H-3 Release = uCi/cc H-3 x cfm x 28320cc/cf Rate 60 min/sec
{ H-3 release (Ci/s) = #ilCVcc H-3 x building exhaust flow rate (cfm) x 28320 cc/cf x 1/60 min/s)
- For a WGDT Rupture accident, the default H-3 release is 2500 Ci over one hr or 6.94E+05 gCVs for 1 hr
- 7) IF tntium accident also involves noble gases, THEN CALCULATE Total TEDE as follows:
TEDE (rem) Tritium TEDE (rem) Total TEDE (rem) 6 mile IF. Calculating 6 mi THYROID CDE Doses I
- 1. CIRCLE the Thyroid CDE FACTOR (rem/h per gCi/m), based on the stability class and release level.
NOTE: USE ground level for all cases except for BFN stack.
A I l C D E F G Ground 2.OE-06 l 2.5E-06 l 5.0E-06 I 2.2E-05 l 4.7E-05 l 1.2E-04 I 2.7E-04 Stack 2.OE-06 2.5E-06 4.5E-06 7.7E-06 3.5E-06 2.3E-07 1.OE-1 1
- 2. OBTAIN the 1-131 release rate from B.4 (gCi/s).
- 3. CALCULATE Thyroid CDE Dose as follows:
______________X ________X ____/__
1-131 release rate (Cis) Thyroid CDE FACTOR duration (hrs) wind speed Thyroid CDE (item 2) (item 1) (item C.1) (m/s) 5 mile (item A.1)
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 22 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES___l Appendix F Page 8 of 8 Manual Method for Assessing Airborne Releases E 3 lml m
[-G. Summary of Results Site Boundary TEDE Rates
- 1. Total TEDE Rate (item C.7) rem/h
For 0.62 mi TEDE dose rate 2 REP Emergencv Class 1E-04 rem/h NOUE 1E-02 rem/h ALERT IE-01 rem/h SAE 1E+00 rem/h GE Site Boundary Thyroid CDE Dose Rate
- 3. CDE Dose Rate (section D.3) rem/h
For 0.62 mi CDE dose rate 2 REP Emergencv Class NA NA 0.5 rem/h SAE 5 rem/h GE 5 Mile TEDE
[ Mile Thyroid CDE
Emergency Class
- 8. Circle the most restrictive REP class from items 2 and 4: NOUE Alert SAE GE END OF MANUAL Calculated by: (initial / date) /
ASSESSMENT DATA VERIFICATION Verified by: (initial / date) /
- Revision
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 23 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES APPENDIX G COMPARISON OF MEASURED FIELD DATA TO DOSE PROJECTION MODELS Ratio = PROJECTED centerline dose rate MEASURED centerline dose rate Unlikely - Field Team Plant but release possible not on unidentified?
NOBLE GAS RATIO Not unusual centerline?
I I I I 11I I 0.001 0.01 0.1 0.2 0.5 I I
10 I I 60 100 1000 Base actions on Base Attempt to verify projections by verified field actions traversing the plume. Ask for measurements on reevaluation and level of projections confidence on plant release data. Must be very confident that field data is correct to override projections based on measured release data.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 24 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES. ___
APPENDIX H WATERDOSE WATERDOSE Evaluation of Liquid Release to the River Code Run
- 1. Log on to WATERDOSE. DOUBLE CLICK on the "CECC VA)C icon. PRESS return until prompted for username. ENTER usemame (WATERDOSE) and password (CECC).
- 2. OBTAIN the following information for input to WATERDOSE and FOLLOW code prompts.
NOTE: TYPE CTRL Z any time to exit or re-start program.
- a. Determine Release Point:: El Diffuser El Shoreline
- b. Length of release: Hours
- c. Volume of release: (ft3)
- d. Release Mix (nuclides and concentrations)
Nuclide Concentration (CI/ml)
- 3. RUN the WATERDOSE code using the available (or default below) information to obtain an estimate of the dose impact. (If the computer code is not operational, the dose calculation methodology contained in Appendix I can be used.)
BFN-33000 cfs SQN - 29000 cfs WBN - 2700 cfs INOTE: TYPE CTRL Z any time to exit or re-start program.
- 4. OBTAIN the State Update Form (SUF). The Preparer and Verifier shall initial and date the results.
- 5. TRANSMIT (by spooling through the computer) the SUF to the TSC and State if approval to do so has been given by the RAC/RAM.
- 6. TYPE CTRL Z any time to exit or re-start program.
APPENDIX I Page 1 of 5 MANUALl Manual Evaluation of Liquid Releases to the River RIVER
- 2. Release Point: ODiffuser OShoreline
- 3. Release Time: Start End _ _ Eastern oCentral
- 4. Release Volume (V) ft3 (1 gal = 0.134 f 3)
- 5. Calculation of Hazard Index (HI):
Nuclide Concentration Dose Hazard (gCiml) (rem/day per gCi/ml) Index (rem/day)
C DF- Table 3 HI=C
- DF Total Hazard Index
- 6. Riverflow at the plant flt3 /s (cfs). This can be obtained from the ICS (for SQN and WBN) or River Operations. If flow data is not available use the following default values:
BFN-33000 cfs SQN - 29000 cfs WBN - 2700 cfs
- 7. Calculate the downstream dose rate to hypothetical Individual at first downstream Public Water Supply and then other locations of interest. Refer to Appendix J.
(table 2) (table 1) (item 5) (item 4)
Location Arrival Time Dilution Factor Hazard Index Release Dose Rate TRM Hours (1/ft 3) (rem/day) Volume (ft3) (rem/day)
________ D Hi V D*H*V
- 8. Record the applicable data on the State Update Form in CECC EPIP-1 and distribute.
Comments:
END OF MANUAL Calculated by: (initial / date) : -
ASSESSMENT DATAVERIFICATION Calculated by: (initial /date)
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 26 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES APPENDIX I Page 2 of 5 TABLE I Manual Evaluation of Liquid Releases to the River DILUTION FACTOR RELATIVE CONCENTRATION FACTORS-PER CUBIC FOOT RELEASED L (D)
BROWNS FERRY NUCLEAR PLANT SHRPFLIINF PI FASE flIPPI Iq;Fp PIPP PFI F=AQr Tennessee Plant Side Opposite River Shoreline Shoreline Centerline Shoreline Mile (TRM) D D D D 294.00 Plant _ _
293.00 2.8E-08 .OOE+00 1.4E-08 .OOE+00 292.00 1.4E-08 .OOE+00 6.9E-09 .OOE+00 291.00 9.2E-09 .OOE+00 4.6E-09 1.2E-36 290.00 6.9E-09 .OOE+00 3.4E-09 8.2E-30 289.00 5.5E-09 .OOE+00 2.8E-09 1.OE-25 288.00 4.6E-09 .OOE+00 2.3E-09 5.1 E-23 287.00 3.9E-09 .OOE+00 2.OE-09 4.3E-21 286.00 3.4E-09 .OOE+00 1.7E-09 1.2E-1 9 285.00 3.1 E-09 .OOE+00 1.5E-09 1.5E-18 284.00 2.8E-09 1.8E-42 1.4E-09 I 1.2E-1 7 283.00 2.5E-09 1 .8E-39 1.3E-09 6.2E-1 7 282.00 2.3E-09 5.8E-37 1.1E-09 2.4E-16 281.00 2.1E-09 7.4E-35 1.1E-09 7.7E-16 280.00 2.OE-09 4.8E-33 9.8E-10 2.1E-15 279.00 1.8E-09 1.8E-31 9.2E-10 4.8E-15 278.00 1.7E-09 4.1E-30 8.6E-10 1.OE-14 274.90 Downstream Dam SEQUOYAH NUCLEAR PLANT
_QHtR10I INF RFI FbAF flIFFI IQPD 0IOF 0F1 PA=F Plant Side Opposite Shoreline Shoreline Centerline Shoreline TRM D D D D 484.50 Plant 484.00 3.5E-08 5.9E-34 1.8E-08 1.1E-14 483.00 1.4E-08 5.3E-1 9 7.2E-09 3.OE-1 I 482.00 9.1 E-09 3.2E-15 4.6E-09 1.9E-10 481.00 6.6E-09 1.5E-13 3.3E-09 3.9E-10 480.00 5.2E-09 1.4E-12 2.6E-09 5.6E-10 479.00 4.3E-09 5.4E-12 2.2E-09 6.8E-10 478.00 3.7E-09 1.4E-11 1.8E-09 7.6E-10 477.00 3.2E-09 2.7E-1 I 1.6E-09 8.2E-10 476.00 2.8E-09 4.5E-11 1.4E-09 8.4E-10 475.00 2.5E-09 6.6E-1 I 1.3E-09 8.6E-10 474.00 2.3E-09 9.OE-1 1 1.2E-09 8.6E-10 473.00 2.1 E-09 1.2E-10 1.1 E-09 8.6E-10 472.00 1.9E-09 1.4E-10 1.OE-09 8.5E-10 471.00 1.8E-09 1.7E-10 9.8E-10 8.3E-10 471.00 Downstream Dam
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 27 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES APPENDIX I Page 3 of 5 TABLE I Manual Evaluation of Liquid Releases to the River DILUTION FACTOR RELATIVE CONCENTRATION FACTORS-PER CUBIC FOOT RELEASED (D)
WATTS BAR NUCLEAR PLANT SHORELINE RELEASE DIFFUSER PIPE RELEASE Plant Side Opposite Shoreline Shoreline Centerline Shoreline TRM D D D D 528.00 Plant _
527.00 3.7E-08 1.2E-20 1.8E-08 2.3E-11 526.00 1.8E-08 1.5E-14 9.1 E-09 4.1 E-09 525.00 1.2E-08 1.3E-12 6.1 E-09 1.OE-09 524.00 9.1E-09 1.2E-11 4.6E-09 1.4E-09 523.00 7.3E-09 4.OE-1 1 3.7E-09 1.7E-09 522.00 6.1E-09 9.0E-11 3.1E-09 1.8E-09 521.00 5.2E-09 1.6E-10 2.7E-09 1.8E-09 520.00 4.6E-09 2.3E-1 0 2.4E-09 1.8E-09 519.00 4.1 E-09 3.1E-10 2.2E-09 1.8E-09 518.00 3.7E-09 3.8E-10 2.OE-09 1.8E-09 517.00 3.3E-09 4.62-10 1.9E-09 1.7E-09 516.00 3.0E-09 5.2E-10 1.8E-09 1.7E-09 515.00 2.8E-09 5.8E-10 1.7E-09 1.6E-09 514.00 2.6E-09 6.4E-10 1.6E-09 1.6E-09 513.00 2.4E-09 6.8E-10 1.6E-09 - 1.5E-09 512.00 2.3E-09 7.22-10 1.5-01.5E-09 511.00 2.2E-09 7.6E-10 1.5E-09 1.4E-09 510.00 2.0E-09 7.92-10 1.4E-09 1.4E-09 510.00 1.9E-09 8.2E-10 1.4E-09 1.4E-09 508.00 1.8E-09 8.4E-10 1.3E-09 1.3E-09 507.00 1.8E-09 8.6E-10 1.3E-09 1.3E-09 506.00 1.7E-09 8.7E-10 1.3E-09 1.3E-09 505.00 1.6E-09 8.8E-10 1.2E-09 1.2E-09 504.00 1.5E-09 8.9E-10 1.2E-09 1.2E-09 503.00 1.5E-09 9.0E-1 0 1.2E-09 1.2E-09 502.00 1.4E-09 9.1E-10 1.2E-09 1.2E-09 501.00 1.4E-09 9.1E-10 1.1E-09 1.1E-09 500.00 1.3E-09 9.1E-10 1.1 E-09 1.1 E-09 471.00 Downstream Dam
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 28 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES ___
APPENDIX I Page 4 of TABLE 2 Manual Evaluation of Liquid Releases to the River ARRIVAL TIME APPROXIMATE TRAVEL TIME TO MAXIMUM CONCENTRATION - HOURS (HRS)
BROWNS FERRY NUCLEAR PLANT TRM RIVER FLOW IN CUBIC FEET/SECOND 25000 30000 33000 35000 37000 39000 294.00 Plant 292.00 25 21 19 18 17 16 290.00 49 41 37 35 33 31 288.00 74 62 56 53 50 47 286.00 99 82 75 71 67 63 284.00 124 103 92 88 84 80 282.00 148 124 112 106 100 94 280.00 173 144 131 124 117 110 278.00 198 165 150 141 134 128 276.00 222 185 169 159 150 142 274.90 Downstream Dam SEQUOYAH NUCLEAR PLANT
___ _ 21000 25000 29000 30000 33000 484.50 Plant 483.00 5 4 4 3 3 481.00 12 10 8 8 7 479.00 18 15 13 13 12 477.00 25 21 18 17 16 475.00 32 26 23 22 20 473.00 38 32 28 27 24 471.00 45 38 32 31 28 471.00 Downstream Dam WATTS BAR NUCLEAR PLANT 19000 20000 25000 30000 35000 528.00 Plant 526.00 5 4 3 3 3 524.00 10 9 7 6 6 522.00 15 14 11 g 9 520.00 20 19 15 12 12 518.00 25 24 19 16 15 516.00 30 29 23 19 18 514.00 35 33 27 22 21 512.00 40 38 30 25 24 510.00 45 43 34 29 28 508.00 50 48 38 32 31 506.00 56 53 42 35 34 504.00 61 58 46 38 37 502.00 66 62 50 41 40 500.00 71 67 54 45 43 471.00 Downstream Dam
> 3
~~~~~~~~~~~~~~~~~TA APPENDIX I Page 5 of 5 DOSE Manual Evaluation of Liquid Releases to the River FACTORS Critical Inaestion Dose Rate Factors (DF)
(Derived from Regulatory Guide 1.109)
Rem/day per i.Ci/ml Nuclide Dose Factor Organ 1 Age2 Nudide Dose Factor Organ 1 Age2 H-3 0.28 TS C Ru-103 43.2 GIT A C-14 16.9 B C Ru-105 58.9 GIT C Na-24 8.12 TB C Ru-106 356 GIT A P-32 1155 B C Ag-110m 121 GIT A Cr-51 1.34 GIT A Te-125m 21.8 K A Mn-54 28 GIT A Te-127m 55 K A Mn-56 67.8 GIT C Te-127 25.8 GIT C Fe-55 16.1 B C Te-129m 96 K C Fe-59 68 GIT A Te-129 20.4 GIT I Co-58 30 GIT A Te-131 168 GIT A Co-60 80.4 GIT A Te-131 6.4 GIT I Ni-63 753 B C Te-132 154 GIT A NI-65 35.8 GIT C 1-130 1332 THY I Cu-64 14.2 GIT A 1-131 12500 THY I Zn-65 51.1 L C 1-132 142 THY I Zn-69 12.33 GIT I 1-133 2980 THY I Br-83 0.24 TB C 1-134 37.4 THY I Br-84 0.28 TB C 1-135 584 THY I Br-85 0.013 TB C Cs-134 538 L C Rb-86 153 L I Cs-136 90.4 L C Rb-88 0.45 L I Cs-137 438 B I Rb-89 0.26 L I Cs-138 1.13 GIT I Sr-89 1850 B C Ba-1 39 50.2 GIT I Sr-90 23800 B C Ba-140 116 B C Sr-91 74.2 GIT C Ba-141 7.27 GIT I Sr-92 239 GIT C Ba-142 1.06 GIT I Y-90 204 GIT A La-140 185 GIT A Y-91m 2.43 GIT I La-142 46.3 GIT C Y-91 155 GIT A Ce-141 48.4 GIT A Y-92 146 GIT C Ce-143 91.2 GIT A Y-93 238 GIT C Ce-144 330 GIT A Zr-95 61.8 GIT A Pr-143 80.6 GIT A Zr-97 210 GIT A Pr-144 4.44 GIT I Nb-95 42 GIT A Nd-147 69.8 GIT A Mo-99 39.8 K C W-187 56.4 GIT A To-99m 1.44 GIT C Np-239 48 GIT A 1.THY = thyroid, GIT = Gastrointestinal Tract, K Kidney, L = Liver, TB = Total Body, B = Bone
- 2. A = Adult, C = Child, I = Infant
Jo APPENDIX J Page 1 of 3 BFN - PUBLIC AND INDUSTRIAL SURFACE WATER SUPPLIES Type of County-State Plant Name Water Source Water SuvvlV Notification Advise State or Local Authorities 10-Mile Radius listed in the REND Limestone-Alabama Browns Ferry Nuclear Plant Tennessee River Industrial Lawrence-Alabama W. Morgan, E. Lawrence Tennessee River Municipal Lawrence-Alabama Water Authority Champion International Tennessee River Industrial &
25-Mile Radius (Courtland Plant) Potable State of Alabama Joe Wheeler State Park Tennessee River Municipal Lawrence-Alabama TVA-Wheeler Dam' Tennessee River Industrial 50-Mile Radius Lauderdale-Alabama Florence City-Wilson Plant Tennessee River Municipal Colbert-Alabama Reynolds Metals Company Tennessee River Industrial Colbert-Alabama Muscle Shoals Tennessee River Municipal Fleet Hollow Embayment Colbert-Alabama TVA ERL Fleet Hollow Embayment Industrial &
Potable Colbert-Alabama TVA-Wilson Dam Tennessee River Industrial Colbert-Alabama Occidental Chemical Company Tennessee River Industrial Colbert-Alabama Sheffield Tennessee River Municipal Colbert-Alabama Sheffield Police Colbert-Alabama TVA Colbert Fossil Plant Tennessee River Industrial Colbert-Alabama Cherokee Water Works & Gas Tennessee River Municipal Colbert-Alabama Cherokee Police (Day)
Colbert-Alabama Cherokee Police (Night)
Colbert-Alabama Laroche Industries Tennessee River Industrial
'Potable water obtained from East Lauderdale County Water District.
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR Page 31 of 32 PLANT RADIOLOGICAL CECC EPIP-8 Revision 26 EMERGENCIES APPENDIX J Page 2 of 3 SQN - PUBLIC AND INDUSTRIAL SURFACE WATER SUPPLIES Type of County-State Plant Name Water Source Water SuDDIY Notification Advise State or Local Authorities 10-Mile Radius listed in the REND Hamilton-Tennessee Sequoyah Nuclear Plant Tennessee River Industrial Gold Point Marina Tennessee River East Side Utility Tennessee River Industrial Tennessee River Industrial Chickamauga Dam (Power Service Center) Tennessee River Industrial Chickamauga Dam Tennessee River 25-Mile Radius E. I. Dupont Co. Tennessee River Industrial and Potable Tennessee American Water Co. Tennessee River Municipal Rock-Tennessee Mill1 Tennessee River Industrial Vulcan Sand & Gravel' Tennessee River Industrial Signal Mountain Cement' Tennessee River Industrial Medusa Cement Co. Tennessee River Industrial 50-Mile Radius Signal Mountain Cement (Plant) Tennessee River Industrial Marion-Tennessee Signal Mountain Cement (Quarry) Tennessee River Industrial South Pittsburg Tennessee River Municipal Nickajack Dam Tennessee River Industrial Jackson-Alabama Bridgeport Tennessee River Municipal and Spring Jackson-Alabama Bridgeport Police Tennessee River and Spring
'Obtains potable water from Tennessee-American Water Company.
- 2 Obtains potable water supply from Bridgeport - physically removed potable water intake in November 1986.
1, ..
(~.
C ( I'"
DOSE ASSESSMENT STAFF ACTIVITIES DURING NUCLEAR PLANT RADIOLOGICAL EMERGENCIES CECC EPIP-8 Page 32 of 32 Revision 26 7 R~~~~~~~
APPENDIX J Page 3 of 3 WBN - PUBLIC AND INDUSTRIAL SURFACE WATER SUPPLIES ON THE TENNESSEE RIVER Type of County-State Plant Name Water Source Water Supply Notification Advise State or Local Authorities listed in 10-Mile Radius the REND Rhea-Tennessee Watts Bar Fossil & Hydro Plant Tennessee River Industrial 2 ,3 Watts Bar Nuclear Plant Tennessee River *IndustrialS34 25-Mile Radius Rhea-Tennessee City of Dayton Tennessee River Municipal Dayton Police 50-Mile Radius Hamilton-Tennessee TVA Sequoyah Nuclear Plant Tennessee River Industrial East Side Utility Tennessee River Industrial E. I. Dupont Tennessee River Industrial and Potable Chickamauga Dam Tennessee River Industrial Tennessee American Water Co. Tennessee River Municipal Rock-Tennessee Mills Tennessee River Industrial Vulcan Sand and Gravels Tennessee River Industrial Signal Mountain Cements Tennessee River Industrial
'On layby status - water use when activated Is about 445 MGD.
2 Cooling water.
3 Potable water to nuclear plant, steam plant, hydro plant, and resort area, provided through Watts Bar Reservation System (wells).
dCooling water and cooling tower makeup.
5 Obtains potable water supply from Tennessee-American Water Company.
CECC EPIP Coversheet Title CECC EPIP-13 Tennessee REV. 10 Valley Authority
! CENTRAL EMERGENCY CONTROL CENTER TERMINATION AND RECOVERY EMERGENCY PLAN IMPLEMENTING Effective 12ate:
PROCEDURES 71ollo-;
WRITTEN BY: Al Salatka /1 REVIEWED BY: 6LvS, /1o Signature Signature Date PLAN EFFECTIVENESS DETERMINATION:0 ZI ec? a2_4-t Signature Date CONCURRENCES Concurrence Signature Date 0g lM r,u EP Prn f f Planning and Implementation 0 Manage mergency Pre redness D9 M.1nag Radiologial and Chemistry Services 0 7/011l3
_~~ ~ ~ ~ ~ ~ ~ ~ _
CECC-EPIP-1 3 TERMINATION AND RECOVERY REVISION LOG Rev. No. Date Revised Pages 0 3/22/88 All (Formerly IP-16; changed from IPD to EPIP)
I 7/8/88 Page 1 2 12/12/88 All 3 7/13/89 All 4 6/20/90 All-*formerly EPIP-23 (former EPIP-13 transferred to EPIP-14) 5 5/15/92 Pgs. 2 & 3 revised. New coversheet and rev. log added.
All pages issued.
6 9/27/95 All pages revised.
7 10/30/96 Pg. 3 remove reference to Appendix C. Procedure put in new format. All pages issued.
8 7/10/00 Annual review and self-assessment items. All pages issued.
9 3/19/03 Annual review. Add information for NRC Administrative Letter 97-03. All pages issued.
10 7/1/03 The procedure was completely rewritten to provide for a graded transition from termination phase to recovery phase based on the severity of the event. Separate checklists were provided for the termination and recovery phases. A organization chart for the recovery organization was added. Procedure put in new EPIP format. All pages issued.
.1 i TERMINATION AND CECC EPIP-13 Page 1 of 15 RECOVERY Revision 10 Table of Contents 1.0 PURPOSE ....... 2 2.0 SCOPE ...... 2 3.0 RESPONSIBILITIES .............................................. 2 4.0 PROCEDURES .............................................. 3 4.1 Termination ............................ 3 4.2 Recovery Operations ........................... 3 4.2.1 Accident Assessment And Investigation ............................ 4 4.2.2 Recovery Planning And Scheduling ............................ 4 4.2.3 Repair and Restoration Activities ............................ 5 5.0 LOCAL RECOVERY CENTER (LRC) .............................................. 6 6.0 ENVIRONMENTAL SAMPLE COLLECTION AND ANALYSIS ...................................... 6
7.0 REFERENCES
.............................................. 6 8.0 ABBREVIATIONS .............................................. 6 Appendixes A. CECC Director's Termination Checklist 7 B. CECC Director's Recovery Checklist.10 C. NRC Administrative Letter 97-03 .11 D. Recovery Organization............... 16
TERMINATION AND RECOVERY ECCPage 2 of 15 Revision 10 1.0 PURPOSE
- This procedure provides guidance on termination and recovery from an incident for which onsite and offsite emergency centers were activated by the Site Emergency Director and transition from the Emergency Response Organization to the Recovery organization if necessary.
- Termination begins when personnel responsible for the response effort determine that
- conditions are sufficiently stabilized to begin comparing them to pre-established
- decisional criteria. The termination decision and subsequent notification that an event
- no longer constitutes an Operational Emergency establishes the beginning of recovery.
- Recovery is defined as those actions taken, after a plant has been brought to a stable or
- shutdown condition, to return the plant to normal operation. Recovery will begin when
- the emergency response is declared terminated. The level of recovery operations
- depends on the severity of the event. The recovery phase may be implemented in a
- graded approach from one of no recovery actions necessary to a fully implemented
- course of actions. When implemented, the recovery phase continues until the plant and
- any affected areas meet predetermined criteria for the resumption of normal operation or
- use.
- Types of activities conducted during the recovery phase may include (but are not limited
- to):
- Damage assessment
- Environmental consequence assessment
- Long-term protective action determinations
- Plant and/or environmental restoration
- Dissemination of information 2.0 SCOPE
- This procedure applies to the termination of a REP event which required activation of onsite and offsite emergency centers and actions for reentry and recovery activities required to restore the plant to normal operating condition and to provide assistance to state and local organizations.
- 3.0 RESPONSIBILITIES
- 3.1 The Senior Vice President, Nuclear Operations, or his designee will direct the overall recovery effort. If expected to be a long-term process, he may establish a recovery organization to be responsible for continuous direction and control of the recovery operation. This organizational structure would be contingent upon the emergency situation and required actions for recovery.
Staffing of the CECC may remain in whole or in part as necessary. The LRC is also available to provide additional office space near the site to support the recovery operation.
- 3.2 The CECC Director is responsible for coordinating with the Site Emergency Director, NRC, and
- appropriate offsite agencies in determining when to enter the recovery phase. Once that
- decision has been made, the CECC Director will notify the Senior Vice President, Nuclear Operations, or his designee.
- Revision
TERMINATION AND RECOVERY CECC EPIP-13 Page 3 of 15 Revision 10 If the event was associated with an emergency off-site either natural or manmade which impacted the off-site (State and local) emergency response, the NRC regional administrator will inform the affected license when the condition of the off-site emergency preparedness infrastructure can support a safe reactor restart. NRC Administrative Letter 97-03 which provides information for plant restart discussions following natural disasters is provided as Appendix C.
- designees with:
- drafting news releases concerning progress of the recovery operation
- coordinating all news releases with TVA management and State and Federal officials as
- required.
coordinating all press briefings and interviews concerning the incident.
- 3.4 Radiological Assessment Manager (RAM) provides radiological support as necessary.
- 3.5 The Vice President, Engineering and Technical Services, will provide required technical support to the site.
- 3.6 The Manager, Nuclear Fuels, will provide needed technical services to the site. Technical services available Include fuel management and core analysis, core performance, nuclear fuel control and accountability, and startup support.
- 4.0 PROCEDURES
- 4.1 Termination The decision to terminate an incident for which onsite and offsite emergency centers have been activated will be made by the Site Emergency Director after consultation with the plant technical and operations staffs and coordinated with the CECC Director. Proposals for termination of an emergency and entry into recovery will be coordinated with the State and NRC, if appropriate,
- through the CECC. Termination decisions should be based on site-specific EPIP-16 criteria
- and broad-based parameters such as:
- Radiation or hazardous material exposure levels within the affected plant or area(s) are stable or
- decreasing with time.
- The affected plant is in a stable condition, and there Is a high probability that it can be
- maintained in that condition.
- Releases of hazardous material to the environment have ceased or are controlled within
- permissible regulatory limits, and the potential for an uncontrolled release is low.
- All emergency notifications have been completed.
- The Site Emergency Director and CECC Director In consultation with the NRC and appropriate
- offsite agencies do not identify a valid reason to continue operating in the emergency response mode.
- Initial recovery activities have been clearly Identified and prioritized.
- When applicable, a recovery staffing plan has been developed, approved and can be
- implemented.
- Revision
TERMINATION AND CECC EPIP-13 Page4of15 RECOVERY Revision 10 4.2 Recovery Operations Recovery planning and implementation will start with assessment of plant, site, and environmental conditions. There are three general areas of recovery operations: accident assessment and investigation, recovery planning and scheduling, and repair and restoration.
4.2.1 Accident Assessment and Investigation The following type of activities should be considered for accident assessment and investigation:
- Plant management in coordination with TVAN Corporate management, should establish an investigation board to determine the root cause of the event and prepare a formal accident report.
- All documents generated during the emergency response and useful to the accident investigation should be collected and organized.
- Plant technical, operations, and maintenance staffs should assess the condition of the plant including structural integrity, equipment status, hazardous material containment/confinement barriers, and safety systems.
- Provide support, when requested, to federal, state, and local government agencies for assistance with offsite dose assessment and related activities.
4.2.2 Recovery Planning and Schedulinc The following type of activities should be considered for recovery planning and scheduling:
- Notification to persons and agencies involved in the emergency response of the establishment of the Recovery Organization and the name of the person in charge.
- Evaluation of emergency plans to determine if adequate emergency preparedness status can be maintained during degraded plant conditions (e.g., inaccessibility of assembly areas, inoperative emergency/safety instrumentation and equipment, etc.)
- Establishment of specific criteria to be met prior to the resumption of normal operations or facility use.
- Contact with the affected State to coordinate any support required for assessment and recovery of affected offsite areas.
- Preparation of plans for the establishment of safe long-term conditions when the assessment indicates that a plant or affected area cannot be safely returned to normal operation or use.
Entire Page Revised
l TERMINATION AND RECOVERY CECC EPIP-13 l Page 5 of 15 Revision 10
- Identification of required repair and restoration work based on the assessment results.
- Plan for the proper handling and disposal of all hazardous waste generated during recovery activities.
- Establishment of a tracking organization to monitor all assigned tasks, including developing work packages, scheduling activities, and estimating costs.
- Formation of a procedures review group to determine if specialized procedures are required and should be developed and to review and approve all special procedures.
- Continued evaluation of site or facility hazards and contamination levels during estimating exposure to workers.
4.2.3 Repair and Restoration Activities The following type of activities should be considered for repair and restoration activities:
- Ensure that occupation exposure limits are followed in accordance with SPP-5.1, Radiological Controls.
- Ensure that any discharges form recovery activities are controlled within regulatory and environmental compliance limits. If discharges are necessary beyond these limits, ensure all documentation Is prepared, approvals obtained, and notifications made.
- Conduct recovery activities through normal work organizations, practices, limitations, and procedures to the extent practical.
- Replenish, repair, or replace any emergency equipment or consumable materials used during the emergency response.
- Train applicable personnel on changes that occurred as a result of repair, restoration, and accident investigation.
Entire Page Revised
f .,
TERMINATION AND RECOVERY CECC EPIP-13 Page 6of15 Revision 10
- 5.0 LOCAL RECOVERY CENTER (LRC)
- 5.1 The purpose of the LRC is to provide a nearsite facility for TVA recovery management as well as NRC emergency response personnel and other emergency and/or recovery personnel.
- 5.2 The LRC provides adequate space for TVA and others who may locate there to support the site should additional office space near the site become necessary during the recovery phase.
- 5.3 The LRC will provide space for NRC personnel. Adequate supplies, communications, and data necessary for them to carry out appropriate functions is available.
- 6.0 ENVIRONMENTAL SAMPLE COLLECTION AND ANALYSIS
- 6.1 The TVA emergency field monitoring vans will be used to collect appropriate samples. This sample collection will be coordinated with the State. Samples will be divided and delivered to the State and the appropriate TVA laboratory.
- 6.2 Western Area Radiological Laboratory (WARL) will perform (or coordinate performance by
- approved testing facilities) environmental sample analysis. Information concerning the samples will be provided to the State and the RAM.
7.0 REFERENCES
NP Radiological Emergency Plan NRC Administrative Letter 97-03 CECC EPIP
- 8.0 ABBREVIATIONS WARL - Western Area Radiological Laboratory.
NP - Nuclear Power.
LRC - Local Recovery Center.
CECC - Central Emergency Control Center.
SED - Site Emergency Director.
- Revision
TERMINATION AND RECOVERY CECC EPIP-13 Revision 10 APPENDIX A Page 1 of 3 CECC DIRECTOR'S TERMINATION CHECKLIST 0O
- Radiation or hazardous material exposure levels within the affect plant or area(s) PAM:
are stable or decreasing with time.
RAM:
CECC Dir.:
El YES 0 NO Date: Time:
2
- The affected plant is In a stable condition, and is there a high probability that it can PAM:
be maintained in that condition (site-specific EPIP-16 criteria verified by RAM:
CECC Plant Assessment and Radiological Assessment staffs). CECC Dir.:
_ _______ O YES 0 NO Date: ime:
3
- Releases of hazardous material to the environment have ceased or are PAM:
controlled within permissible regulatory limits, and the potential for an RAM:
uncontrolled is release low.
CECC Dir.:
0 YES D NO Date: Time:
4
- All emergency notifications have been completed. CECC Dir.:
lo YES a NO Date: Time:
5
- The Site Emergency Director and CECC Director, in consultation with the NRC and appropriate offsite agencies agree that no valid reason exists to continue operating in the emergency response mode.
CECC Dir.:
DYES E NO
_ __ _ __ _ _ __ ______________________________________ Date: r im e:
Entire Page Revised
TERMINATION AND Pg f1 RECOVERY CECC EPIP-13 R8eof 15 APPENDIX A Page 2 of 3 CECC DIRECTOR'S TERMINATION CHECKLIST 6 0
make appropriate notifications to the affected state.
Date: Time:
____C OYES O NO 7
- The Senior Vice President, Nuclear Operations has been notified of event termination.
aQYES O NO Stop here if no recovery actions are CEOC Dir.:
necessary.
If recovery operations are necessary, continue with Steps 9 & 10 and continue Date: Time:
to CECC EPIP-13, Appendix B.
9 If applicable, Initial recovery activities
~~~~*
have been clearly identified and PAM:
prioritized.
RAM:
CECC Dir.:
0 YES 0 NO Date: Time:
10 .l If applicable, a recovery staffing plan has been developed, approved, and can be PAM:
implemented.
RAM:
CECC Dir.:
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TERMINATION AND l I Page 9 of 15 I ~CEL.~~L.I~tJRECOVERY CE I-3Revision 10 APPENDIX A Page 3 of 3 CECC DIRECTOR'S TERMINATION CHECKLIST EVENT TERMINATION:
The: 0 NOUE El ALERT E SITE AREA EMERGENCY 0 GENERAL EMERGENCY Affecting: BFN U2 El, U3 SON U El, U2 0 WBN Ul [I EAL Designator:
HAS BEEN TERMINATED Event Termination Time: _ Date: _
Call affected State and provide this Information State Notification Time: Date:
CECC Director:
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TERMINATION RECOVERY AND CECC EPIP-13 Pae1 fI PRevision 1 APPENDIX B Page of I CECC DIRECTOR'S RECOVERY CHECKLIST Ai ac io The recovery organization has been established. CECC Dir.:
YES El NO Date: Time:
2 Accident Assessment and Investigation activities have been considered and implemented as determined, based on the severity of the event, including the collection and organization of all documents generated during the emergency response. CECC Dir.:
[ YES El NO Date: rime:
3 EO The affected state agency has been contacted to coordinate any support required for assessment and recovery of affected offsite areas.
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4 a Appropriate Recovery Planning and Scheduling activities have been considered and Implemented as determined, based on the severity of the event.
CECC Dir.:
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5 a The NRC has been contacted as applicable to NRC Administrative Letter 97-03. Refer to Appendix C.
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tI i TERMINATION AND l Page11 OfI l RECOVERY CECC EPIP-13 ~~~~Revision 10 APPENDIX C NRC ADMINISTRATIVE LETTER 97-03 Page 1 of 4 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 March 28,1997 NRC ADMINISTRATIVE LETTER 97-03: PLANT RESTART DISCUSSIONS FOLLOWING NATURAL DISASTERS Addressees All holders of operating licenses or construction permits for nuclear power reactors.
Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this administrative letter to inform addressees about a recently adopted Internal practice. This practice involves coordinating the assessment of offsite recovery and onsite restart activities following a natural disaster (hurricane, tomado, flood, storm, earthquake, etc.) where offsite damage may be substantial or undetermined. This administrative letter does not transmit or imply any new or changed requirements or staff positions. No specific action or written response Is required.
Background
Numerous events have occurred in recent years In which natural disasters have affected power reactor facilities. Most notable of these Is Hurricane Andrew and Its impact on the Turkey Point Station. The licensee for the Turkey Point plant shut the reactors down in anticipation of the storm. Onsite damage from the hurricane was extensive. After that event, the licensee repaired the damage and was ready to restart the plant before the offsite emergency preparedness infrastructure was ready to support the restart. An assessment of offsite conditions and infrastructure prior to restart was necessary to assure emergency preparedness in the event of a subsequent reactor accident.
Events have also occurred in which plants have shut down in anticipation of hurricane damage, which turned out to be minimal. Despite the absence of onsite damage, either some offsite damage occurred that affected the state of offsite emergency preparedness, or some damage occurred offsite such that the state of offsite emergency preparedness could not be determined Immediately.
For these cases, the NRC coordinated with the Federal Emergency Management Agency (FEMA) and the licensees involved to ensure that the restarts occurred after the offsite emergency preparedness Infrastructure could safely support them.
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TERMINATION ANDPage 12 of5 RECOVERY CECC EPIP-13 Revision 10 APPENDIX C NRC ADMINISTRATIVE LETTER 97-03 Page 2 of 4 Discussion Although the overall responsibility for confirming the adequacy of radiological emergency preparedness of commercial nuclear power plants is vested with the NRC, it relies on FEMA's assessment of offsite emergency planning and response activities when carrying out this responsibility.
Section III of the Memorandum of Understanding (MOU) Between FEMA and the NRC, dated June 17, 1993, lists responsibilities for both agencies for cooperating in the recovery from a disaster that affects the offsite emergency preparedness infrastructure surrounding power reactors. FEMA's headquarters (HQ) in Washington, D.C., is responsible for providing findings and determinations to the NRC concerning the adequacy of offsite preparedness in the areas surrounding power reactor sites following a severe natural event.
FEMA HQ bases its assessment on information from State and local governmental authorities, as well as from the affected FEMA regional office and the NRC.
In two recent instances (Hurricane Bertha, July 1996 and Hurricane Fran, September 1996), FEMA HQ chartered special evaluation teams to assess whether the offsite emergency preparedness infrastructure could support the restart of plants that had shut down in anticipation of hurricanes that affected the sites. These teams consisted of FEMA and NRC regional representatives, State and local emergency management representatives, and, in a limited capacity, power reactor licensee personnel. These teams provided assessments to FEMA HQ for its ultimate determinations that offsite emergency preparedness could support plant restart in both cases. The chartering of these special evaluation teams helped ensure a timely assessment of the condition of the offsite infrastructure and was based on experience gained with Hurricane Opal (October 1995) and the Quad Cities tomado (May 1996).
In some cases, a natural disaster may occur where onsite damage is minimal, but offsite damage may be substantial or undetermined. In these cases, the plant may be ready to start up shortly after the event. Communications in these cases between the licensee and NRC, the NRC and FEMA, and FEMA and offsite officials will be aggressive; however, stringent protocols will be observed to ensure that FEMA and the NRC operate within the guidelines of the MOU.
The NRC uses FEMA's determinations to inform power reactor licensees when the condition of the offsite emergency preparedness infrastructure can support a reactor restart. The Office of Nuclear Reactor Regulation (NRR), as well as NRC regional offices, have adopted a communication protocol that links key personnel in the two agencies and the affected licensee organization. An overview of this protocol is attached. Some of the key points of this protocol are:
- 1. NRC regional office personnel maintain close contact with the affected power reactor licensee to determine the state of onsite emergency preparedness and the plans for restart. The NRC regional office communicates this information rapidly to NRR.
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- 2. FEMA regional office personnel maintain close contact with their evaluators in the field, the affected State and local emergency management officials, and the affected NRC regional office to determine the state of offsite emergency preparedness. The FEMA regional office communicates this information rapidly to FEMA HQ.
- 3. The final assessment that offsite emergency preparedness can support a power reactor restart originates from FEMA HQ.
- 4. A single Individual in NRR serves as the point of contact with FEMA HQ to receive this assessment. The individual communicates this Information rapidly to NRR management and the cognizant NRC regional office.
- 5. After the assessment from FEMA is received and discussed with NRR management, the NRC regional administrator Informs the affected licensee that the condition of the offsite emergency preparedness infrastructure can support a safe reactor restart.
The NRC has developed this protocol as a result of discussions with FEMA, as well as lessons learned from Hurricane Andrew and other events. The objective of this protocol is to ensure that aggressive and rapid Information flow occurs between the involved organizations following natural disasters at power reactors. The NRC expects that the use of this protocol will ensure that the determination that the condition of the offsite emergency preparedness infrastructure can support a reactor restart will be made before the licensee is actually ready to restart the reactor plant(s). In the event that the determination is not made before the licensee is ready to restart the plant(s), the NRC will evaluate the need to delay the restart through the issuance of an order or confirmatory action letter. By accomplishing this protocol, the licensee, FEMA, and NRC can provide for safe and rapid restarts of power reactors in the wake of these disasters and assure that the offsite emergency preparedness Infrastructure can function as expected If called upon in an emergency.
This administrative letter requires no specific action or written response.
If you have any questions about this letter, please contact the contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
signed by Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation
Contact:
W. Maier, NRR (301) 415-2926 E-mail: wam@nrc.gov Attachments:
- 1. Information Flow for Restart Considerations Following Natural Disasters at Power Reactors
APPENDIX C NRC ADMINISTRATIVE LETTER 97-03 Page 4 of 4 Attachent AL 97-03 March 28, 1997 iLI Page 1 of
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RECOVERY ECC EPIP-13 evision 10 APPENDIX D Page 1 of 1 RECOVERY ORGANIZATION RadlologicalManager, Cherny Sri and es
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