ML20258A282

Submittal of Changes to Emergency Plan Implementing Procedures

ML20258A282
Person / Time
Site:	Grand Gulf
Issue date:	09/14/2020
From:	Emily Larson Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20258A281	List: GNRO-2020/00026, 3 to Emergency Plan Implementing Procedure 10-S-01-1, Activation of the Emergency Plan Safety Related ML20258A282
References
GNRO-2020/00026
Download: ML20258A282 (321)

Text

Entergy Operations, Inc.

P.O. Box 756 Port Gibson, Mississippi 39150 Eric A. Larson Site Vice President Grand Gulf Nuclear Station Tel: 601-437-7500 10 CFR 50.54(q)

GNRO-2020/00026 September 14, 2020 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Changes to Emergency Plan Implementing Procedures Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29

Dear Sir or Madam:

Entergy Operations, Inc. hereby submits, in accordance with 10 CFR 50.54(q)(5), changes to the following Emergency Plan Implementing Procedures:

10-S-01-1 Revision 131 Activation of the Emergency Plan 10-S-01-1 Revision 132 Activation of the Emergency Plan There are no commitments contained in this submittal. If you have any questions or need additional information, please contact Mike Lewis at 601-437-2166.

Sincerely, EAL/saw

Attachment:

10-S-01-1 Revision 131 Activation of the Emergency Plan 10-S-01-1 Revision 132 Activation of the Emergency Plan 50.54(q) Revision 131 Activation of the Emergency Plan 50.54(q) Revision 132 Activation of the Emergency Plan

GNRO-2020/00026 Page 2 of 2 cc: NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

GNRO-2020/00026 Attachment To GNRO-2020/00026 10-S-01-1 Revision 131 Activation of the Emergency Plan 10-S-01-1 Revision 132 Activation of the Emergency Plan

Attachment 2 Page 1 of 3 10CFR50.54(Q)(3) Screenina Procedure/Document Number: Activation of the Revision: 131 Emergency Plan/ 10-S-01-1 Equipment/Facility/Other: GGNS

Title:

Implement NRC Approved Rev 6 EALs Part I. Description of Activity Being Reviewed (This is generally changes to the emergency plan, EALs, EAL bases, etc. - refer to Section 3.0 Step 6):

Implementation of EAL Revision 6.

By application dated April 27, 2018, as supplemented by letter dated October 10, 2018, Entergy Operations, Inc. (the licensee) requested US. Nuclear Regulatory Commission (NRC, the Commission) approval of changes to revise the emergency plan for Grand Gulf Nuclear Station, Unit 1 (GGNS). The proposed amendment would revise the current emergency action level (EAL) scheme to one based on Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012.

Revision 6 of NEI 99-01 was endorsed by the NRC in a letter dated March 28, 2013.

The supplemental letter dated October 10, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed. and did not change the NRG staffs proposed no significant hazards consideration determination as published in the Federal Register on June 5, 2018 (83 FR 26104).

~YES NO Part II. Activity Previously Reviewed? 50.54(q)(3) Continue to Evaluation is next part Is this activity fully bounded by an NRC approved 10CFR50.90 submittal or NOT required.

Alert and Notification System Design Report? Enter justification below and If YES, identify bounding source document number/approval reference and complete Part ensure the basis for concluding the source document fully bounds the VI.

proposed change is documented below:

Justification: The Changes are fully bounded in GNRI 2019-00029 - NRC Approval of Rev 6 EALs.

D Bounding document attached (optional)

Part Ill. Applicability of Other Regulatory Change Control Processes Check if any other regulatory change processes control the proposed activity. (Refer to EN-Ll-100)

APPLICABILITY CONCLUSION D If there are no other controlling change processes, continue the 10CFR50.54(q)(3) Screening.

D One or more controlling change processes are selected, however, some portion of the activity involves the emergency plan or affects the implementation of the emergency plan: continue the 10CFR50.54(q)(3) Screening for that portion of the activity. Identify the applicable controlling change processes below.

D One or more controlling change processes are selected and fully bounds all aspects of the activity. 10CFRs0.54(q)(3)

Evaluation Is NOT required. Identify controlling change processes below and complete Part VI.

EN-EP-305 ROOS

Attachment 2 Page 2 of 3 1 0CFR50.54(Q)(3) Screening Procedure/Document Number: Activation of the Revision: 131 Emergency Plan/ 10-S-01-1 Equipment/Facility/Other: GGNS

Title:

Implement NRC Approved Rev 6 EALs CONTROLLING CHANGE PROCESSES Part IV. Editorial Change YES ONO 50.54(q)(3) Continue to Evaluation is next part Is this activity an editorial or typographical change such as formatting, paragraph NOT required.

numbering, spelling, or punctuation that does not change intent? Enter justification Justification: and continue to next part or complete Part VI as applicable.

Part V. Emergency Planning Element/Function Screen (Associated 10CFR50.47(b) planning standard function identified in brackets) Does this activity affect any of the following, including program elements from NU REG-0654/FEMA REP-1 Section II?

1. Responsibility for emergency response is assigned. (1]

2. The response organization has the staff to respond and to augment staff on a continuing basis (24/7 staffing) in accordance with the emergency plan. [1)

3. The process ensures that on shift emergency response responsibilities are staffed and assigned. (2]

4. The process for timely augmentation of onshift staff is established and maintained. [2)

5. Arrangements for requesting and using off site assistance have been made. f3] n

6. State and local staff can be accommodated at the EOF in accordance with the emergency plan. [3]

7. A standard scheme of emergency classification and action levels is in use. [4]

8. Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes after declaration of an emergency and providing follow-up notifications. [5]

9. Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (5]

10. The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. [5]

11. Systems are established for prompt communication among principal emergency response organizations. [6]

12. Systems are established for prompt communication to emergency response personnel. [6)

13. Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). [7)

14. Coordinated dissemination of public information during emergencies is established. [7)

EN-EP-3O5 ROOS

Attachment 2 Page 3 of 3 10CFR50.54(Q)(3) Screenina Procedure/Document Number: Activation of the Revision: 131 Emergency Plan/ 10-S-01-1 Equipment/Facility/Other: GGNS

Title:

Implement NRC Approved Rev 6 EALs

15. Adequate facilities are maintained to support emergency response. [8] D

16. Adequate equipment is maintained to support emergency response. [8) D

17. Methods, systems, and equipment for assessment of radioactive releases are in use. [9] D

18. A range of public PARs is available for implementation during emergencies. [10] D

19. Evacuation time estimates for the population located in the plume exposure pathway EPZ are D available to support the formulation of PARs and have been provided to State and local governmental authorities. [1 OJ

20. A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.[1 OJ

21. The resources for controlling radiological exposures for emergency workers are established. [11 J

22. Arrangements are made for medical services for contaminated, injured individuals. [12)

23. Plans for recovery and reentry are developed. [13]

24. A drill and exercise program (including radiological, medical, health physics and other program areas) is established. [14)

25. Drills, exercises, and training evolutions that provide performance opportuniti~s to develop, maintain, and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses. [14]

26. Identified weaknesses are corrected. [14]

27. Training is provided to emergency responders. [15)

28. Responsibility for emergency plan development and review is established. [16)

29. Planners responsible for emergency plan development and maintenance are properly trained. [16)

APPLICABILITY CONCLUSION

J If no Part V criteria are checked, a 10CFR50.54(q)(3) Evaluation is NOT required; document the basis for conclusion h~ 1~** __ ..., - o- ... \II

J If any Part V criteria are checked, complete Part VI and perform a 10CFR50.54(q)(3) Evaluation.

BASIS FOR CONCLUSION Part VI. Signatures:

Preparer Name (Print) Date:

Dennis M. Coulter /" 1+*1, (Optional) Reviewer Name (Print) Reviewer Signature Date:

ReviewerName(Print) ~~ *_ 'Si ature,, Date:

James J. Lewis Nuclear EP Project Manager Approver Name (Print)

Michael L. Lewis V /J1r ver Signature

/

Date:

Emergency Planning Manager or designee  ?/A,/ 5 1 /2-z./2020 I

EN-EP-305 ROOS

Attachment 2 Page 1 of 4 1 0CFR50.54(Q)(3) Screening Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan Part I. Description of Activity Being Reviewed (This is generally changes to the emergency plan, EALs, EAL bases, etc. - refer to Section 3.0 Step 6):

The following are changes to the 10-S-01-1 Change# Page Description of Change 1 Flow Charts Changes made for ease of use. Added background color of referred section to notes.

1.a Cold S/D Changed Cold 5/D category background color from salmon Category to light blue 1.b CG UE Block Added reference under C6 UE block and 58 UE Block with S8 UE Block yellow background: "Refer to Category H for potential UE due to a Hazardous Event" 1.c S4 Alert Added note with green background under 54 Alert block:

Block "Refer to Table F-1 for potential escalation due to degraded fuel clad barrier."

1.d S5 Alert Added note with green background under 55 Alert block:

Block "Refer to Table F-1 for potential escalation due to degraded RCS barrier."

1.e H2 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a seismic event" 1.f H3 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a hazardous event" 1.g H4 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a fire" 1.h H2 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a seismic event" 1.i H3 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a hazardous event."

1.j H4 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a fire."

2 Flow Charts Added the computer monitoring points for release pathways in Table A-1 3 Flow Charts Added page number reference for EAL basis document for each EAL 4 Flow Charts Added procedure number and revision date 5 Flow Charts Added controlled document number box.

6 Flow Charts Changed EAL CA1.1 RPV water level (Level 2) from <-42 in.

to <-41.6 in.

EN-EP-3O5 ROOS

Attachment 2 Page 2 of 4 10CFR50.54(Q)(3) ScreeninQ Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan 7 Attachment 2 Changed EAL CA1.1 RPV water level (Level 2) from <-42 in.

CA1.1 Pages to <-41.6 in. Changed reference to GGNS TRM Table 80 and 81 TR3.3.4.2-1 8 Attachment 2 Corrected second reference from 1 to 2.

CNB2 Page 161 9 Attachment 2 Shaded the EAL boxes of all EALs for ease of locating the EALs 10 Attachment 2 Removed numerical sub-contents. Removed "7.0 page 1 Table Attachments" from TOC and replaced with EAL Listing of Contents and page numbers.

11 Attachment 3 Corrected degree symbol for the Main Turbine Lube Oil Page 12 temperature range.

12 Procedure Revised Attachment numbering. Attachment 1 is now Page 4 and the flow charts with the remaining attachments Attachment increased by one. Changed the attachment numbers Pages from Roman numerals to Arabic numbers.

YES 0NO Part II. Activity Previously Reviewed? 50.54(q)(3) Continue to Evaluation is next part Is this activity fully bounded by an NRC approved 10CFR50.90 submittal or NOT required.

Alert and Notification System Design Report? Enter justification below and If YES, identify bounding source document number/approval reference and complete Part ensure the basis for concluding the source document fully bounds the VI.

proposed change is documented below:

Justification:

D Bounding document attached (optional)

Part Ill. Applicability of Other Regulatory Change Control Processes Check if any other regulatory change processes control the proposed activity.(Refer to EN-Ll-100)

None APPLICABILITY CONCLUSION 0 If there are no other controlling change processes, continue the 10CFR50.54(q)(3) Screening.

D One or more controlling change processes are selected; however, some portion of the activity involves the emergency plan or affects the implementation of the emergency plan; continue the 10CFR50.54(q)(3) Screening for that portion of the activity. Identify the applicable controlling change processes below.

D One or more controlling change processes are selected and fully bounds all aspects of the activity. 10CFR50.54(q)(3)

Evaluation is NOT required. Identify controlling change processes below and complete Part VI.

EN-EP-305 R008

Attachment 2 Page 3 of 4 10CFR50.54(Q)(3) Screenina Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan CONTROLLING CHANGE PROCESSES 10 CFR 50.54(q)

Part IV. Editorial Change i:8lYES NO 50.54(q)(3) Continue to Evaluation is next part Is this activity an editorial or typographical change such as formatting, paragraph NOT required.

numbering, spelling, or punctuation that does not change intent? Enter justification Justification: Changes 1-5 and 8-12 are editorial. These changes either add and continue amplifying information or correct typographical errors in the procedure. to next part or complete Part "No" is checked because this revision contains other changes that are not VI as applicable.

editorial. Changes 6 and 7 are material and are being screened.

Part V. Emergency Planning Element/Function Screen (Associated 10CFR50.47(b) planning standard function identified in brackets) Does this activity affect any of the following, including program elements from NU REG-0654/FEMA REP-1 Section II?

1. Responsibility for emergency response is assigned. [1]

2. The response organization has the staff to respond and to augment staff on a continuing basis (24/7 staffing) in accordance with the emergency plan. [1]

3. The process ensures that on shift emergency response responsibilities are staffed and assigned. [2]

4. The process for timely augmentation of OnShift staff is established and maintained. [2]

5. Arrangements for requesting and using off site assistance have been made. [3]

6. State and local staff can be accommodated at the EOF in accordance with the emergency plan. [3]

7. A standard scheme of emergency classification and action levels is in use. [4] [2]

8. Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes after declaration of an emergency and providing follow-up notifications. [5]

9. Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. [5]

10. The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and Notification Systems for Nuclear Power Plants or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. [5]

11. Systems are established for prompt communication among principal emergency response organizations. [6]

12. Systems are established for prompt communication to emergency response personnel. [6]

13. Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). [7]

14. Coordinated dissemination of public information during emergencies is established. [7]

15. Adequate facilities are maintained to support emergency response. [8]

16. Adequate equipment is maintained to support emergency response. [8]

17. Methods, systems, and equipment for assessment of radioactive releases are in use. [9]

EN-EP-3O5 ROOS

Attachment 2 Page 4 of 4 10CFR50.54(Q)(3) Screening Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan

18. A range of public PARs is available for implementation during emergencies. (1 0]

19. Evacuation time estimates for the population located in the plume exposure pathway EPZ are available to support the formulation of PARs and have been provided to State and local governmental authorities. [1 0]

20. A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.[1 0]

21. The resources for controlling radiological exposures for emergency workers are established. [11]

22. Arrangements are made for medical services for contaminated, injured individuals. [12]

23. Plans for recovery and reentry are developed. (13]

24. A drill and exercise program (including radiological, medical, health physics and other program areas) is established. [14]

25. Drills, exercises, and training evolutions that provide performance opportunities to develop, maintain, and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses. [14]

26. Identified weaknesses are corrected. [14]

27. Training is provided to emergency responders. [15]

28. Responsibility for emergency plan development and review is established. [16]

29. Planners responsible for emergency plan development and maintenance are properly trained. [16]

APPLICABILITY CONCLUSION

J If no Part V criteria are checked, a 10CFR50.54(q)(3) Evaluation is NOT required; document the basis for conclusion below and complete Part VI.

~ If any Part V criteria are checked, complete Part VI and perform a 10CFR50.54(q)(3) Evaluation.

BASIS FOR CONCLUSION Changes 1-5 and 8-12 are editorial.

Screening has determined that Planning Standard 10 CFR 50.47(b)(4) is potentially affected by modifying the Level 2 RPV water level value.

A 10 CFR 50.54(q)(3) Evaluation will be performed on changes 6 and 7 to determine whether the effectiveness of the emerqencv plan is reduced, and prior NRC approval is required.

Part VI. Signatures:

~ ~

Preparer Name (Print) / / "

Dennis M. Coulter

• C) z., I B. Z,.O Date:

(Optional) Reviewer Name (Print) Reviewer Signature Date:

,---)/7

/ /

Reviews, Name ( P ~ e v i e w e , Sigoatoce Date:

JamesJ.Lewis ~~

Nuclear EP Project

~

er , .2*--/,f-dd Approver Name (Print) Approver Signature Date:

Michael L. Lewis Emergency Planning Manager or designee

/J/_ l1w1~  ;;/, ~ /~J..>,? 0 EN-EP-3O5 ROOS

Attachment 3 Page 1 of 3 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan Part I. Description of Proposed Change:

The following are changes to the 10-S-01-1 Change# Page Description of Change 6 Flow Charts Changed EAL CAl.1 RPV water level (Level 2) from <-42 in.

to <-41.6 in.

7 Attachment 2 Changed EAL CAl.1 RPV water level (Level 2) from <-42 in.

CA1.1 Page to <-41.6 in. Changed reference to GGNS TRM Table 80 TR3.3.4.2-1 Part II. Description and Review of Licensing Basis Affected by the Proposed Change:

Manually reviewed sections 1, 4 and Appendix A of the Grand Gulf Emergency Plan. No additional relevant or affected plan content was identified. A historical search of previous revisions is not applicable in this case because this revision implements NRC approval of the NEI 99-01 Rev 6 EALs for GGNS for the first time.

EAL content in previous revisions of the emergency plan is no longer relevant Part Ill. Describe How the Proposed Change Complies with Relevant Emergency Preparedness Regulation(s) and Previous Commitment(s) Made to the NRC:

10 CFR 50.47(b)(4} -A Standard Scheme of Emergency Classification and Action Levels is in Use -

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. The GGNS EAL scheme as modified for RPV water level in EAL CA1 .1 still implements a standard scheme with a procedural and technical basis as expressed in the TRM and annunciator panel procedures.

Previous NRC Commitments - Commitments made in AECM 89-0202, GNRO 97-00080, and the GGNS On-Shift Staffing Analysis Report were reviewed. Commitments in AECM 89-0202 were related to tracking due dates of Emergency Planning training to ensure ERO personnel remain qualified while actively on the ERO. The commitment in GNRO 97-00080 is related to clearly defining responsibilities of the on-shift radiological assessment function. Revising the Level 2 RPV water level value does not affect these commitments. All current NRC commitments continue to be maintained.

Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change:

10 CFR 50.47(b)(4} -A Standard Scheme of Emergency Classification and Action Levels is in Use - A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

10 CFR 50 Aggendix E Section IV.B 1-2 reguires a licensee to develog EALs that are used for notification and garticigation of State and Local Agencies.

NUREG 0654 Section 11.J.7 reguires licensees to establish a mechanism for recommending grotective actions to the aggrogriate State and local authorities.

EN-EP-305 R008

Attachment 3 Page 2 of 3 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:

Evaluation of Revising EAL CA 1.1 -

By application dated April 27, 2018, as supplemented by letter dated October 10, 2018, Entergy Operations, Inc. (the licensee) requested US. Nuclear Regulatory Commission (NRC, the Commission) approval of changes to revise the emergency plan for Grand Gulf Nuclear Station, Unit 1 (GGNS). The proposed amendment would revise the current emergency action level (EAL) scheme to one based on Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012. Revision 6 of NEI 99-01 was endorsed by the NRC in a letter dated March 28, 2013.

The supplemental letter dated October 10, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed. and did not change the NRC staff proposed no significant hazards consideration determination as published in the Federal Register on June 5, 2018 (83 FR 26104).

The initial submittal and supplemental letter used a value of RPV level <-42 in. for CA 1.1. The submitted comparison matrix states that -42 in. is the Level 2 actuation setpoint for HPCS and RCIC as justification for the value used. Whereas this value can be found in the referenced CA1 .1 alann response instruction 04 02-1H13-P601 as an annunciator description, the alann response instruction also cites RPV water level of

-41.6 in. as a possible cause of the annunciation. In addition, the TRM lists RPV level of

-41.6 in.as the trip setpoint for HPCS and RCIC, which better aligns with the justification provided in the submitted documents than a value of -42 in. The TRM also indicates that RPV level of -41.6 in. is the Level 2 value, not -42 in. GGNS operators are very familiar with use of the -41.6 in. RPV level as it is currently used in the NEI 99-01 Rev 5-based scheme at the station.

The effectiveness of the emergency plan is not reduced because use of the -41.6 in. RPV level better aligns with the justification provided for the EAL value related to Level 2 actuation for HPCS and RCIC, provides the same value for RPV Level 2 that is described in the TRM and provides a value familiar to the operators in EAL application.

Conclusion -

The proposed change to EAL CA 1.1 continue to meet the planning standards outlined in 10 CFR 50.47(b)(4). This change does not represent a reduction in the effectiveness to the emergency plan and can be incorporated without prior NRC approval.

EN-EP-3O5 ROOS

Attachment 3 Page 3 of 3 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: 10-S-01-1 I Revision: 132 Equipment/Facility/Other: GGNS

Title:

Activation of the Emergency Plan Part VI. Evaluation Conclusion Answer the following questions about the proposed change.

I. Does the proposed change comply with 10CFR50.47(b) and 10CFR50 l'.8:JYES ONO Appendix E?

1. Does the proposed change maintain the effectiveness of the emergency plan l'.8:JYES ONO (i.e., no reduction in effectiveness)?

3. Does the proposed change constitute an emergency action level scheme YES t8J NO change?

If questions 1 or 2 are answered NO, or question 3 answered YES, reject the proposed change, modify the proposed change and perform a new evaluation or obtain prior NRC approval under provisions of 10CFR50.90. If questions 1 and 2 are answered YES, and question 3 answered NO, implement applicable change process(es). Refer to Section 6.7 Step 8.

Part VII. Signatures Preparer Name (Print) Date:

Dennis M. Coulter Z*/8*2.. 0 (Optional) Reviewer Name (Prin Date:

Reviewer Name (Print) Date:

James J. Lewis Nuclear EP Project Manager Approver Name (Print) / ' _,./ .- ;V)Approver Signature Date:

Michael L. Lewis L/,., ~;/

Emergency Planning Manager or designee /}; U J EN-EP-305 ROOS

GNRO-2020/00026 Attachment To GNRO-2020/00026 50.54(q) Revision 131 Activation of the Emergency Plan 50.54(q) Revision 132 Activation of the Emergency Plan

Attachment 4 Page 1 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 131 SECTION 1- PROCEDURE INFORMATION Procedure

Title:

Activation of the Emergency Plan PRGGN/CR No.: PRGGN 2020-00091 Database File ID: - -N/A TCN Number(s): _N_/A Level of Use Changed Procedure Title Changed Procedure Number Changed Revision Change Details/Justifications:

(Pages 3 and 4, Revision Change Details - Continued may be used, if needed.)

Revision 131 implements the NRC approved Revision 6 EAL Scheme per GNRI 2019-00029.

This revision is a complete re-write of the EALs and revision bars will not be used.

SECTION 2 - WORKFLOW PUP: Total Rewrite for Human Factoring P-FIN: Maintains Existing Format New Procedure Minor Revision IZI Major Revision SECTION 3 - REVIEWS Review Type: Print Name Date Procedure Writer: D.M. Coulter 1-14-20 Technical Review: J.J. Lewis 2-17-20 CDR Review:

• Wes Marshall 2-7-20 Validation Review:

• Tre Reeves 2-6-20 2 nd Review:

• N/A

• Reviews not required may be marked N/A 01-S-02-9 R010

Attachment 4 Page 2 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 131 SECTION 4 - REVIEW/EVALUATION Reviews: (MARK appropriate box, Print Name, Sign, and Date - Signature Required)

PAD Not Required Process Applicability Excluded IS/IST Implementation TCN Incorporation PER EN-Ll-100 or Other Process/Number

~

01-S-02-9 0 PAD Required D.M. Coulter I* 1+*2o PER EN-Ll-100 Print Name Date Additional reviews X List additional reviews performed and attach copy of reviews to PRGGN:

were performed and listed in section IV of 50.54Q the PAD form.

SECTION 5 - SUPERVISOR REVIEWS Change Management Plan Required: 0 N/A Yes PRGGN CA initiated: 0 No Yes PRGGN CA No.:

Remarks:

Change Management Plan is not Required. All users of the procedure attended Class Room Training GLP-EP-EALRev6, which reviewed the revised Emergency Action Levels OSRC Review Required: 0 No Yes Michael L. Lewis Print Name f/l I@,:,,,"" J./z.8/zo 1 Date SECTION 6 - OSRC REVIEW ~ NIA Section APPROVED REJECTED Meeting No.:

N/A Print Name Signature Date SECTION 7 - PROCEDURE OWNER or DESIGNEE APPROVAL Michael L Lewis Print Name 111. t-uJ1 Signature z/49/zo I Dafo 01-S-02-9 R010

PLANT OPERATIONS MANUAL Volume 10 10-S-01-1 Section 01 Revision: 131 Date: 2-28-20 REFERENCE USE EMERGENCY PLAN PROCEDURE ACTIVATION OF THE EMERGENCY PLAN SAFETY RELATED Prepared: Dennis M. Coulter Reviewed: James J. Lewis Technical Approved: Michael L. Lewis_______

Manager, Emergency Planning List of Effective Pages:

Pages 1-18 Attachments 1-6 EPP 01-02 (Flowchart): GGNS Rev.[x}

List of TCNs Incorporated:

Revision TCN 1-4 None 5 1,2 6 3 7 4 8 5,6 9 None 10 7,8 11 None 12 9 13 10 14,15 None 16 11 17 None 18 12 19-21 None 22 13,14 23 None 100 15 101 None 102 16 103-120 17 121 018 122-131 None J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: i Emergency Plan Does this directive contain Tech Spec Triggers? ( ) YES (X) NO REQUIREMENTS CROSS-REFERENCE LIST Requirement Implemented by Directive Directive Paragraph Number Name Paragraph Number That Implements Requirement GGNS Emer Plan 2.4.S3 1.1.1 GGNS Emer Plan 3.1.S3 6.1.2 GGNS Emer Plan 3.3.S1 2.1.2 GGNS Emer Plan 3.3.S3 & S4 2.4 GGNS Emer Plan 3.3.S5 2.1.2 GGNS Emer Plan 3.3.S7.b 6.1.3 GGNS Emer Plan 4.1.S3 6.1.6, 6.3 GGNS Emer Plan 4.1.S13 6.1.3 (Note)

GGNS Emer Plan 4.1.4.S3 6.1.6.k(1)

GGNS Emer Plan 4.1 S9 Attachment I GGNS Emer Plan 5.4.S6 6.1.6.j(1),j(2),j(3)

GGNS Emer Plan 5.4.27 2.2 GGNS Emer Plan 5.4.22 2.3 GGNS Emer Plan 6.1.2.S1 & S2 6.1.6.d (Note), g (Note), 6.1.7 GGNS Emer Plan 6.2.2.S1 & S2 6.1.2 (Note)

GGNS Emer Plan 6.2.2.S3 & S4 2.1.1, 6.1.2 S2 GGNS Emer Plan 6.2.4.S2 2.1.2 GGNS Emer Plan 6.2.4.S7 & S8 2.4 GGNS Emer Plan 6.2.4.S11 & S12 6.1.6.j(1)

GGNS Emer Plan 6.2.4.S13 6.1.6.j(2) & (3)

GGNS Emer Plan 6.3.1.S1 & S2 6.2.1 GGNS Emer Plan 6.3.2 6.2.1a GGNS Emer Plan 6.3.3, 6.3.4 6.2.1b GGNS Emer Plan 6.5.1.a.1 6.1.6.g, 6.1.6.h GGNS Emer Plan 6.5.1.b 6.1.6 GGNS Emer Plan 6.5.1.b.S5,S6,& S8 6.1.6.k(1)

GGNS Emer Plan 7.3.1 S13-17 6.1.7.j GGNS Emer Plan 7.3.3 S9-12 6.1.7.j GGNS Emer Plan 7.5.3.a.2.e 5.19, 6.1.6.f GGNS Emer Plan 7.5.3.a.3.c 6.1.6.e GGNS Emer Plan 9.0.S6 Attachment III (Note)

GGNS Emer Plan 9.3.S7 3.6 GGNS Emer Plan Table 4-1 Attachment I ANSI N18.7 5.3.9.2.S1

• GNRO-97/00113.97-15-02 Item 2 6.2.1.a(3)

CNRO-05/00044(35518) Attachment I, HS1.1, HA1.1, HU1.1 NUREG - 1022 3.1.1 6.1.5 IER L1-13-10

• LO-GLO-2017-00028

• Covered by directive as a whole or by various paragraphs of the directive.

Current Revision Statement Revision 131:

Implement NEI Rev 6 EALs J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 1 Emergency Plan Table of Contents Page 1.0 PURPOSE AND DISCUSSION 2 2.0 RESPONSIBILITIES 2

3.0 REFERENCES

3 4.0 ATTACHMENTS 4 5.0 DEFINITIONS 4 6.0 DETAILS 9 6.1 Activation of Emergency Plan 9 6.2 Supplemental Actions 16 6.3 Upgrading Emergency Classifications 17 6.4 Terminating Emergency 17 6.5 Records and Reports 17 6.6 EP form EPP 01-02 (Flow Chart) Revision Process 18 J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 2 Emergency Plan 1.0 PURPOSE AND DISCUSSION 1.1 Purpose 1.1.1 This procedure provides guidance to:

a. Classify an emergency according to severity.

b. Assign responsibilities for emergency actions.

c. Establish lines of authority and communication.

d. Initiate emergency actions to safeguard the public and plant personnel.

e. Upgrade or terminate emergency classification when severity of event changes.

1.2 Discussion 1.2.1 Whenever plant conditions are identified that meet the Emergency Action Level Criteria in Attachment 1 or EPP 01-02 (Flowchart),

this emergency plan procedure shall be implemented.

1.2.2 Emergency Plan section 6.1.2 allows suspension of normal Emergency Plan actions for Security Emergencies. This is permitted because of the potential risk to personnel safety which a security emergency may present. An armed attack against the plant is a unique security emergency that is expected to be an extremely fast moving event and present an immediate and serious threat to human life. Section 6.1.8 of this procedure contains special Emergency Plan actions for this unique event.

1.3 Changes required for implementation of 1994 TSIP were incorporated in Revision 100. For historical reference this statement should not be deleted.

2.0 RESPONSIBILITIES 2.1 Shift Manager - Is responsible for determining if emergency declaration is required.

2.1.1 If an Emergency Action Level (EAL) is reached or exceeded, the Shift Manager shall:

a. Classify the emergency and make the appropriate declaration if required.

b. Take action to ensure safe operation of plant and protection of plant personnel, the general public, and plant equipment.

c. Perform assessment actions.

d. Perform any other emergency actions as appropriate.

2.1.2 The Shift Manager assumes the role of Emergency Director upon initial classification of an emergency, and resumes normal Control Room duties when relieved by the EOF Emergency Director.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 3 Emergency Plan 2.2 Operations Coordinator - The Operations Coordinator reports to the TSC Manager. Responsibilities include:

2.2.1 Coordinate TSC efforts in determining the nature and extent of emergencies pertaining to equipment and plant facilities in support of Control Room actions.

2.2.2 Assist the EPM in evaluating changes in event classification.

2.2.3 Ensure the Control Room, TSC, and EOF is informed of significant changes in event status.

2.2.4 Coordinate operations activities outside of the Control Room with the TSC Manager and OSC Manager.

2.3 Security Coordinator - The Security Coordinator is located in the Incident Command Post and reports to the Emergency Plant Manager. Responsibilities include:

2.3.1 Overall coordination of the offsite assistance for the security related response.

2.3.2 Designated NIMS (National Incident Management System) Liaison between the Incident Command Post and Site Organization.

2.4 EOF Emergency Director - Is responsible for:

2.4.1 Reporting to the site to assume the duties of Emergency Director upon notification of an Alert or higher classification.

2.4.2 Assuming the duties of Emergency Director after the EOF is declared operational.

2.4.3 Reporting to the site to assume duties of Emergency Director upon notification of an Unusual Event if he deems it necessary.

2.4.4 Evaluating the accident conditions and verifying that the correct emergency classification has been made.

3.0 REFERENCES

3.1 NRC Memorandum dated July 11, 1994 concerning "Branch Position on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1".

3.2 GGNS Emergency Plan 3.3 10-S-01-6, Notification of Offsite Agencies and Plant On-Call Emergency Personnel.

3.4 10-S-01-11, Evacuation of Onsite Personnel 3.5 10-S-01-12, Radiological Assessment and Protective Action Recommendations.

3.6 10-S-01-22, Recovery 3.7 10-S-01-23, Reentry 3.8 EN-EP-609, Emergency Operations Facility (EOF) Operations 3.9 EN-EP-610 TECHNICAL SUPPORT CENTER (TSC) OPERATIONS 3.10 05-1-02-VI-4, Security Threat ONEP J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 4 Emergency Plan 3.10 NUREG - 1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73 3.11 10-S-01-38, EAL Contingency Planning.

3.12 IER L1-13-10, Nuclear Accident at Fukushima Daiichi Nuclear Plant l 4.0 ATTACHMENTS 4.1 Attachment 1 - EAL Wall Charts 4.2 Attachment 2 - EAL Technical Bases 4.3 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases 4.4 Attachment 4 - Guidelines to Terminate Emergency 4.5 Attachment 5 - Event Termination Checklist 4.6 Attachment 6 - Rapidly Progressing Severe Accident Determination 5.0 DEFINITIONS NOTE Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS).

These words are defined terms that have specific meanings as used in the EAL and EAL Basis document. The definitions of these terms are provided in Attachment 2.

5.1 Alert - Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

5.2 Assessment Action - Actions taken during or after an accident to obtain and process information necessary to make decisions to implement specific emergency measures.

5.3 Bomb - refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

5.4 CAS - Central Alarm Station 5.5 CDE (Thyroid) (Committed Dose Equivalent) - The radiation dose to the adult thyroid gland due to radioiodines over a fifty year period following inhalation or ingestion.

5.6 Civil Disturbance - is a group of 10 or more persons violently protesting station operations or activities at the site.

5.7 Confinement Boundary - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC).

5.8 Containment Closure - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 5 Emergency Plan Containment Closure is established when either Primary or Secondary Containment integrity is established .

5.9 DLR - Dosimeter of legal Record 5.10 Decay Heat Removal - As long as plant parameters can be controlled within the limits of 05-S-01-EP-2, adequate core cooling and control of decay heat removal exists. Loss of the SSW Basins (Ultimate Heat Sink) capability to remove heat does not by itself cause loss of control of decay heat removal.

5.11 Downwind - An area located beyond a fixed point in the same direction the wind is blowing. The area covers three sectors, the sector containing the plume centerline, and the two adjacent sectors. If the plume is on a sector line, four sectors are used until the three sector criteria can be identified.

5.12 Emergency - A sudden, urgent, usually unforeseen occurrence or occasion requiring immediate action. It may result from accidental causes, natural causes, or malicious man-made actions. There are four classes of emergencies considered: Unusual Event, Alert, Site Area Emergency, and General Emergency.

5.13 Emergency Action Levels (EALs) - A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level.

5.14 Emergency Classification - One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)

Alert Site Area Emergency (SAE)

General Emergency (GE) 5.15 Emergency Director - An individual designated onsite having the authority and responsibility to initiate the Emergency Plan and coordinate efforts to reduce the consequences of the event and bring it under control 5.16 Emergency Operations Facility (EOF) - A near-site emergency center from which the offsite emergency support activities are controlled 5.17 Emergency Planning Zone (EPZ) - Areas designated for which planning is provided to assure that prompt and effective action is initiated to protect the public in the event of an emergency 5.18 EPP - Emergency Plan Procedure 5.19 ERDS - Emergency Response Data System. A near real-time data link from the GGNS Balance of Plant computer to the NRC Operations Center. This system monitors specific data and is activated by the Shift Manager no later than one hour after an ALERT (or higher) declaration.

5.20 ESC - Energy Services Center 5.21 Exclusion Area - Area surrounding the plant, owned by the licensee, in which the licensee has the authority to determine all activities including exclusion or removal of personnel and/or property.

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 6 Emergency Plan 5.22 Explosion - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.)

should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

5.23 Extortion - is an attempt to cause an action at the station by threat of force.

5.24 Fire - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

5.25 General Emergency - Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

5.26 Hostage - is a person(s) held as leverage against the station to ensure that demands will be met by the station.

5.27 Hostile Action - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the plant staff to achieve an end. This includes attack by air, land or water using guns, explosives projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the plant. Non-terrorism based EALs should be used to address such activities, (e.g. violent acts between individuals in the security owner controlled area).

5.28 Hostile Force - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, causing destruction.

5.29 Immediately Dangerous to Life and Health (IDLH) - a condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

5.30 KI - Potassium Iodide 5.31 LOCA - Loss of Coolant Accident 5.32 Lower Flammability Limit (LFL) - the minimum concentration of a combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and a gaseous oxidizer.

5.33 Monitor and Prepare - A type of precautionary action intended to advise the public within the EPZ that a serious emergency at the nuclear power plant exits and that it should monitor the situation and prepare for the possibility of evacuation, SIP, or other protective actions. Further, if an evacuation is underway, officials should ask individuals who are not involved in the evacuation to remain off the roadways to allow those who are instructed to evacuate to do so.

5.34 Normal Plant Operations - activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviations from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERTIONS.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 7 Emergency Plan 5.35 Offsite - For accountability purposes, any area outside the GGNS protected area 5.36 OMT - Offsite Monitoring Team 5.37 Onsite - For accountability purposes, the area within the GGNS protected area 5.38 Operations Support Center (OSC) - Location from which onsite non-Control Room activities are staged and implemented 5.39 PA - Public Address System 5.40 PAG - Protective Action Guide 5.41 PAR - Protective Action Recommendation 5.42 Power Block - The power block consists of the following areas; Unit I & II Turbine Building, Unit I & II Auxiliary Building, Control Building, Radwaste Building, Water Treatment Building, Circulating Water Pump House, Diesel Generator Rooms and Standby Service Water Basins.

5.43 Protected Area - is an area which normally encompasses all controlled areas within the security protected area fence.

5.44 Rapidly Progressing Severe Accident - A significant reactor event with immediate or near-immediate offsite consequences that is intended by the regulator to be easily recognizable. If the decision maker is not sure whether or not such an event is occurring based on the criteria in Attachment 6, then they should assume that a rapidly progressing severe accident is not in progress.

5.45 Sabotage - is deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may NOT meet the definition of SABOTGE until this determination is made by security supervision.

5.46 Safe Shutdown Equipment - Safe Shutdown Equipment - The minimum required systems are as follows: A minimum of six (6) main steam safety relief valves, which can be operated from the Remote Shutdown Panel, Residual Heat Removal Systems A and B (Suppression Pool Cooling, Alternate Shutdown Cooling, and LPCI modes), Standby Service Water Systems A and B, Standby Diesel Generators A and B, ECCS Rooms HVAC, ESF Switchgear Rooms HVAC, Standby Service Water Pump House HVAC, Diesel Generator Rooms HVAC, Remote Shutdown Panel System, Portions of electrical distribution systems (L11, L21, L51, R20, and R21), required to support the above systems 5.47 SAP - Severe Accident Procedure 5.48 SAS - Secondary Alarm Station 5.49 Short Duration - For GGNS Radiological Assessment purposes a release that is expected to last less than one-hour.

5.50 Significant Transient - is an UNPLANNED event involving one or more of the following: (1) automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Trip, (4)

Safety Injection Activation, or (5) thermal power oscillations >10%.

5.51 Site Area Emergency - Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 8 Emergency Plan result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

5.52 Site Security Code Yellow - A security posture based on an identified threat where a confirmed adversary has committed a hostile action or is displaying an obvious sign of hostile intent inside or outside the SOCA.

5.53 Site Security Code Red - A security posture based on a serious identified threat where the protected area is breached or is being breached by an adversary with resources to commit radiological sabotage.

5.54 SSW Standby Service Water 5.55 Strike Action - is a work stoppage within the Protected Area by a body of workers to enforce compliance with demands made on Management. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONs.

5.56 TEDE (Total Effective Dose Equivalent) - Sum of the EDE and CEDE to nonpregnant adults from exposure and intake during an emergency situation.

5.57 TSC - Technical Support Center 5.58 Unplanned - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

5.59 Unusual Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

5.60 Valid - an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

5.61 Visible Damage - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

5.62 Vital Area- Areas within the Protected Area that house safety-related equipment. The failure or destruction of this equipment could directly or indirectly endanger the public health and safety by exposure to radiation.

The following areas are considered Vital Areas: Auxiliary Building (including Containment), Control Building (including Control Room Complex),

Diesel Generator Building, Inverter Room (166 elevation Turbine Building),

SSW Pump and Valve rooms.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 9 Emergency Plan 6.0 DETAILS 6.1 Activation of Emergency Plan 6.1.1 Any person having knowledge of abnormal plant conditions should notify the Control Room Supervisor/Shift Manager.

6.1.2 The Control Room Supervisor/Shift Manager, when notified of abnormal plant conditions, should refer to Attachment 1 or EPP 01-02 (Flowchart) to determine if an emergency action level has been reached. Additional clarifying information for each emergency action level can be found in Attachment 2. If an emergency action level has been reached, the emergency plan shall be implemented.

Grand Gulf Nuclear Station maintains the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and shall promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level.

NOTE The Control Room Supervisor/Shift Manager is responsible for determining if the declaration of an emergency is required.

If a declaration is required, he is responsible for activating the emergency plan.

6.1.3 The assessment, classification, and declaration of an emergency condition is expected to be completed within 15 minutes after the availability of indications (i.e. plant instrumentation, plant alarms, computer displays, or incoming verbal reports) to plant operators that an EAL has been exceeded.

a. The 15-minute criterion is not to be construed as a grace period to restore plant conditions to avoid declaring the event.

b. The emergency declaration should be made promptly without waiting for the 15 minute period to elapse once the EAL is recognized as being exceeded.

c. For EALs that specify duration of the off-normal condition, such as fire lasting 15 minutes, loss of power for 15 minutes, etc.:

(1) The Emergency Director shall make the declaration at the first available opportunity when the time has elapsed (not after an additional 15 minutes)

(2) The declaration should be made before the EAL is met (time duration has elapsed) when the Emergency Director has information that the off-normal condition will not be corrected within the specified time duration J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 10 Emergency Plan 6.1.4 Whenever there is doubt as to the classification of the emergency condition or if more than one EAL is reached, the more conservative classification shall be used.

NOTE When EALs are observed in conjunction with plant or equipment status due to planned maintenance or testing activities, an emergency condition may or may not exist and the situation must be evaluated on a case-by-case basis.

6.1.5 In the event that Emergency Action Level (EAL) indicators or instrumentation is lost for any reason, the Shift Manager or Emergency Director should refer to reference 3.11 which identifies the plant instruments that are designated indicators of EAL entry criteria and analyzes each primary instrument for backup/alternative indications allowed by the EAL, instrument redundancy, and correlation between EAL requirements by Operating Mode versus related Technical Specifications/Technical Requirements Manual Allowed Outage Time and required action.

6.1.6 IF an event or condition existed which met or exceed an Emergency Action Level but an emergency was not declared and the basis for the emergency classification no longer exists at the time of the discovery (rapidly concluded event, missed classification, or misclassified event), THEN PERFORM the following actions:

a. IF ANY of the following conditions exists, THEN CLASSIFY the event and MAKE required notifications to NRC and Offsite Agencies (using an Emergency Notification Form):

(1) The event is a rapidly concluded event and the proper classification is higher than an Unusual Event.

(2) The event caused damage to plant or it cannot be determined if plant damage occurred.

b. IF either of the following conditions are true, THEN no classification of event is required. Notify the NRC within one hour of the discovery of the event, as required by NUREG-1022, in accordance with Administrative procedure 01-S-06-5.

(1) The event is a rapidly concluded event meeting the EAL classification not higher than Unusual Event, is no longer meeting the EAL action level.

(2) The EAL threshold was not recognized at the time of occurrence, but was identified well after the condition occurred (e.g. as result of post event review or record reviews), the condition no longer exists, and there is no further damage to the plant.

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 11 Emergency Plan 6.1.7 Once an emergency classification is declared, the following actions are taken by the Shift Manager/Emergency Director:

NOTE After becoming aware that an emergency condition exists, the Shift Manager/Emergency Director's first priorities are:

• Take actions to ensure safety of plant personnel and general public.

• Take actions to ensure safe operation of plant.

a. In the event of Site Security Code Yellow or Red, go to section 6.1.8 and do not return to this section until the event is over.

NOTE Emergency Director's Checklist (EPP Form 01-1) is to be used by the Control Room Shift Manager as acting Emergency Director making the required notifications and activations after an Emergency Action Level (EAL) is classified and declared. If the Shift Manager / Emergency Director activates the Emergency Response Organization (ERO) the ERO Emergency Director will assume the Emergency Director duties from the Shift Manager after the Emergency Operations Facility (EOF) is declared operational and a formal status turnover is complete as described in EN-EP-609 Attachment 9.1 Emergency Director section 1.2 A through D.

b. Initiate Emergency Director's Checklist (EPP Form 01-1).

c. Announce to Control Room personnel that you are the Emergency Director.

NOTE The NRC shall be notified of the declaration of the emergency IMMEDIATELY AFTER THE NOTIFICATION OF THE STATE AND LOCAL AGENCIES and not later than one hour after the emergency declaration.

d. Designate an individual as communicator to perform the initial notification in accordance with 10-S-01-6. The Shift Manager shall ensure that the primary or secondary state and local agencies are notified within 15 minutes of an emergency declaration or reclassification.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 12 Emergency Plan 6.1.7 (Cont.)

NOTE In the event of Security emergencies, each Security related incident should be evaluated. Only those support groups and facilities which are needed should be activated, regardless of the emergency classification, so as to minimize the risk to personnel. Utilization of the ERO call tree rather than GGNS Computer Notification System may be required to inform responders of emergency situation and prevent manning of unneeded facilities.

e. Activate and verify activation of the GGNS Computer Notification per 10-S-01-6.

f. Activate ERDS within one hour of an Alert or higher declaration Per 10-S-01-6.

g. Announce nature and classification of event:

NOTE For security emergencies, inform all personnel to take immediate cover.

Man only those emergency facilities which are necessary and that dont pose a risk to personnel.

(1) Over Plant PA System or phone #6426.

(2) Over Site Paging (#7929).

h. If an evacuation of affected areas of the plant is required, perform in accordance with 10-S-01-11.

i. Implement plant operating procedures and emergency plan procedures as required to perform emergency corrective and assessment actions.

j. Activate Emergency Response Organization as follows:

(1) If Unusual Event has been declared, no activation of facilities is required unless the Emergency Director feels there is a reasonable possibility of escalation of emergency to a higher classification.

(2) If an Alert or higher classification has been declared, the entire Emergency Organization, all emergency facilities, must be activated.(See Note 6.1.7.g)

(3) If a Site Area Emergency or General Emergency has been declared, a Site Evacuation should be seriously considered.

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Activation of the No.: 10-S-01-1 Revision: 131 Page: 13 Emergency Plan 6.1.7.j (Cont.)

(4) Once activated, the TSC, OSC and EOF shall become operational as soon as possible (without delay). When facility staffing can be accomplished with onsite personnel, it is the goal to become operational within 45 minutes. Otherwise offsite personnel shall provide shift augmentation in 75 minutes and be fully operational in 90 minutes.

k. If an Alert, Site Area Emergency, or General Emergency has been declared, determine offsite doses in accordance with 10-S-01-12.

(1) Protective actions shall be recommended to State and Local Agencies upon declaration of a General Emergency as follows. Sheltering may be recommended instead of evacuation in accordance with reference 10-S-01-12:

Condition Protective Action Recommendation RPSA PAR EVACUATE: 0-2 mile radius and EVACUATE: 2-10 miles downwind Rapidly Progressing and Severe Accident MONITOR: Remainder of 10 mile Emergency AND Planning PREPARE: Zone (EPZ) and Consider use of Potassium Iodide in accordance with State Plans (Standard PAR)

EVACUATE: 2 Miles All Sectors and General Emergency Declared EVACUATE: 5 Miles in Downwind Sectors NOTE: After the initial and issuance of the Standard PAR refer to MONITOR Remainder of 10 Mile Emergency 10-S-01-12 prior to AND Planning Zone (EPZ) with the extending the standard PREPARE: exception of areas previously PAR into additional evacuated.

sectors.

and Consider use of Potassium Iodide in accordance with State Plans J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

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Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 14 Emergency Plan 6.1.7.k (Cont.)

Condition Protective Action Recommendation EXTENDED PAR EVACUATE: 2 Miles All Sectors General Emergency Declared and And EVACUATE: 10 Miles in Downwind Sectors Dose Projection or and Field Measurement at 5 miles MONITOR Remainder of 10 Mile Emergency corresponds to AND Planning Zone (EPZ) with the 1 Rem TEDE PREPARE: exception of areas previously evacuated.

Or and 5 Rem Thyroid.CDE Consider use of Potassium Iodide in accordance with State Plans Standard PAR Required SHELTER PAR And Shelter In Place: 2 mile radius Containment vent of one hour or less with and expected dose < 1rem TEDE and < 5rem CDE Shelter in Place: 5 miles downwind at Site Boundary and area cannot be and evacuated before venting. Monitor and Prepare: Remainder of EPZ

l. Designate shift personnel to perform emergency corrective and assessment actions.

CAUTION The following guidance is only to be used for events involving an armed attack against the plant. Do not use these instructions for any other event.

An armed attack against the plant is a unique security emergency that is expected to be an extremely fast moving event and presents an immediate and serious threat to human life. It is imperative for personnel to take cover immediately to minimize loss of life.

Normal activation of the Emergency Plan is inappropriate for this Event.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 15 Emergency Plan 6.1.8 Activation of the Emergency Plan during an armed attack against the plant.

NOTE Emergency Director's Checklist (EPP Form 01-1) AND 05-1-02-VI-4 Security Threat ONEP shall be used by the Control Room Shift Manager as acting Emergency Director (SM/ED) in the case of a declared CODE RED or CODE YELLOW in the case of an Armed Attack as noted in the Caution Statement above. It is used BEFORE the Emergency Response Organization (ERO) is activated. The ERO Emergency Director will assume the Emergency Director duties from the Shift Manager after the BACKUP Emergency Operations Facility (EOF) at Baxter Wilson Steam Station is declared operational and a formal status turnover is complete as described in EN-EP-609 Attachment 9.1 Emergency Director section 1.2 A through D.

a. Initiate Emergency Directors Checklist (EPP Form 01-01) (CR)

AND 05-1-02-VI-4 Security Threat ONEP (Off-Normal Emergency Procedure)

b. Announce to the Control Room that you are the Emergency Director.

c. Make an announcement over the Plant PA System (or phone

1. 6426) and over Site Paging phone # 7929 to the affect that there is a Site Security Code (Yellow or Red) in affect and all personnel are to take cover immediately until further notice.

d. Take actions as directed by 05-1-02-VI-4, Security Threat ONEP.

e. Do not announce the classification of the event over the Plant PA System.

f. Do not activate any onsite or offsite Emergency Response Facilities. (OSC, TSC, EOF, ENMC, EIC)

g. Activate the GGNS Computer Notification System in accordance with 10-S-01-6.

h. Do not order a site evacuation.

i. Designate an individual as communicator to perform notification in accordance with 10-S-01-6, if available.

j. Ensure Emergency Notification Forms include Protective Action Recommendations in accordance with section 6.1.6.k.(1) of this procedure.

k. If the event occurs during off hours man the backup EOF (Baxter Wilson) with personnel that are not onsite in accordance with reference 3.3.

CAUTION Prior to directing anyone to report to ERFs or any other movements around the site after a security event you must coordinate with Security to ensure personnel do not interfere with security post-event activities.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 16 Emergency Plan 6.1.8 (cont.)

l. When notified that the security threat is over, return to step 6.1.7.b.

6.2 Supplemental Actions 6.2.1 Continuous assessment is necessary to effectively coordinate and direct emergency response. In any emergency situation, attention must be paid to parameters that may indicate a possible worsening of conditions (i.e., radioactive releases).

a. If an Alert condition is declared, the following assessment actions are required:

(1) Increased surveillance of applicable in-plant instrumentation.

(2) Visual observation of affected plant area.

(3) Onsite and offsite radiological monitoring if a release has taken place or is suspected.

(4) Determination of offsite doses if applicable.

b. In addition to the above, a Site Area Emergency or General Emergency would require these additional assessment actions.

(1) Monitor meteorological data.

(2) Dispatch offsite radiological monitoring teams downwind of the release in conjunction with state radiological monitoring efforts.

(3) Assess onsite and offsite radiation doses. (TEDE and Thyroid CDE).

6.2.2 Emergency Director should ensure that periodic announcements are made over the plant PA (or #6426) and site PA (#7929) concerning:

a. Nature and location of event.

b. Required personnel actions.

c. Any other information necessary.

6.2.3 The Emergency Director (while in the Control Room) logs all information in the Shift Manager/Control Room Operator Log as necessary for event reconstruction.

6.2.4 The Emergency Director (while in the EOF) may delegate to the EOF Log Keeper the responsibility for logging all information relative to the emergency (for event reconstruction).

6.2.5 Upon declaring the EOF operational, the following activities should be transferred to the ED as soon as possible:

a. Notifications to offsite agencies

b. Offsite radiological and environmental surveys

c. Protective action recommendations to offsite agencies

d. Classification of the emergency J:\ADM_SRVS/TECH_PUB/REVISION/10/10-S-01-1.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 17 Emergency Plan 6.2.6 If extended emergency operations are necessary, the Emergency Director should authorize preparation of an emergency organization shift schedule to support 24-hour emergency operation.

6.3 Upgrading Emergency Classifications 6.3.1 If conditions worsen, refer to Attachment 1 or EPP 01-02 (Flowchart) to determine if the emergency classification requires upgrading. Additional clarifying information for each emergency action level can be found in Attachment 2. If the classification is upgraded, ensure the following steps are taken:

a. Declare appropriate emergency classification in accordance with Step 6.1.2.

b. Announce nature and classification of event in accordance with Step 6.1.7.g.

c. If an evacuation is required, notify Security if possible and evacuate affected areas in accordance with Step 6.1.7.h.

d. Initiate plant operating procedures and emergency plan procedures as required.

e. Activate additional emergency facilities as necessary in accordance with Step 6.1.7.j.

f. Determine offsite doses in accordance with Step 6.1.7.k.

g. Conduct additional assessment actions as necessary in accordance with Step 6.2.

6.4 Terminating Emergency 6.4.1 Terminating If EALs are no longer met or exceeded, the Emergency Director refers to Attachment 4 to determine whether or not to terminate emergency.

6.4.2 Reentry and Recovery Once the corrective and protective actions taken have established effective control over the situation, the Emergency Director may refer to 10-S-01-23 and 10-S-01-22 to determine if reentry and recovery actions may be initiated.

6.5 Records and Reports 6.5.1 The Manager, Emergency Preparedness is responsible for generating a report on the activation of the Emergency Plan. The report should include the following:

a. Copies of appropriate paperwork generated by the event including: notification forms, checklists, logbooks, survey maps, dose calculations etc.

b. Observations and comments from the personnel involved in the event.

6.5.2 The Manager, Emergency Preparedness is responsible for ensuring that all observations and comments are tracked in accordance with 01-S-10-3.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Activation of the No.: 10-S-01-1 Revision: 131 Page: 18 Emergency Plan 6.6 EP Form EPP 01-02 (Flow Chart) Revision Process 6.6.1 The Manager, Emergency Preparedness is responsible for reviewing all changes to Attachment 1, EPP 01-02 (Flow Chart).

a. If EPP 01-02 (Flow Chart) is changed, before procedure 10-S-01-1 is issued, the Manager, Emergency Preparedness is responsible for verifying the following and that documentation on EP Form EPP 01-03 is complete.

(1) All required training is complete.

(2) Color laminated copies of the revised EPP 01-02 (Flow Chart) are available and stamped with the correct controlled copy number for the following locations:

Control Room (1)

Simulator (1)

TSC (2)

EOF (2)

Back up TSC (1)

Back Up EOF (1)

(3) Non-color copies of the revised EPP 01-02 (Flow Chart) are available and stamped with the correct controlled copy number for all controlled copies of procedure 10-S-01-1, Activation of the Emergency Plan.

b. Once procedure revision for 10-S-01-1, Activation of the Emergency Plan is approved for issue, the distribution of color laminated copies of EPP 01-02 and 10-S-01-1, Activation of the Emergency Plan procedure from document control must be coordinated to ensure all required elements are issued concurrently.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment I EMERGENCY CLASSIFICATIONS EPP 01-02 (Flowchart)

See Flowchart, EPP 01-02 Page 1 of 2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite dose Release of gaseous or liquid radioactivity greater than 2 times the than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE greater than 10 mrem TEDE or 50 mrem thyroid CDE ODCM limits for 60 minutes or longer Prolonged loss of all offsite and all onsite AC power to ESF Loss of all offsite and all onsite AC power to ESF buses Loss of all but one AC power source to ESF buses Loss of all offsite AC power capability to ESF buses buses for 15 minutes or longer for 15 minutes or longer for 15 minutes or longer 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF AG1.1 AS1.1 AA1.1 AU1.1 1 2 3 1 2 3 1 2 3 1 2 3 Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor SG1.1 SS1.1 SA1.1 SU1.1

> column GE for 15 min. (Notes 1, 2, 3, 4) > column SAE for 15 min. (Notes 1, 2, 3, 4) > column ALERT for 15 min. (Notes 1, 2, 3, 4) > column UE for 60 min. (Notes 1, 2, 3) AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 Loss of all offsite and all onsite AC power to DIV I and DIV II Loss of all offsite and all onsite AC power to DIV I and DIV II Loss of all offsite AC power capability, Table S-1, to DIV I AA1.2 ESF 4.16 KV buses ESF 4.16 KV buses for 15 min. KV buses reduced to a single power source for 15 min. and DIV II ESF 4.16 KV buses for 15 min. (Note 1)

AG1.2 AS1.2 AU1.2 AND EITHER: (Note 1) (Note 1) 1 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Sample analysis for a gaseous or liquid release indicates a - Restoration of at least one ESF 4.16 KV bus in < 4 AND

> 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the concentration or release rate > 2 x ODCM limits for 60 min. hours is not likely (Note 1) Any additional single power source failure will result in loss of the SITE BOUNDARY (Note 4) the SITE BOUNDARY (Note 4) SITE BOUNDARY (Note 4) (Notes 1, 2) 1

- RPV water level cannot be restored and maintained all AC power to SAFETY SYSTEMS Loss of > -191 in.

AG1.3 AS1.3 AA1.3 ESF Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or Analysis of a liquid effluent sample indicates a concentration AC Power Table S-1 AC Power Sources Rad or release rate that would result in doses > 10 mrem TEDE or beyond the SITE BOUNDARY: beyond the SITE BOUNDARY:

Effluent - Closed window dose rates > 1000 mR/hr expected to - Closed window dose rates > 100 mR/hr expected to 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for Loss of all vital AC and vital DC power sources for 15 minutes 60 min. of exposure (Notes 1, 2) Offsite continue for 60 min. continue for 60 min. or longer

- Analyses of field survey samples indicate thyroid CDE - Analyses of field survey samples indicate thyroid CDE AA1.4 1 2 3 - ESF Transformer 11

> 5000 mrem for 60 min. of inhalation > 500 mrem for 60 min. of inhalation Field survey results indicate EITHER of the following at or - ESF Transformer 12 (Notes 1, 2) (Notes 1, 2) SG1.2 beyond the SITE BOUNDARY: - ESF Transformer 21

- Closed window dose rates > 10 mR/hr expected to Loss of all offsite and all onsite AC power to DIV I and DIV II continue for 60 min. ESF 4.16 KV buses for 15 min. (Note 1) Onsite 2

AND Loss of all vital DC power for 15 minutes or longer A

- Analyses of field survey samples indicate thyroid CDE - DIV I DG (DG 11)

> 50 mrem for 60 min. of inhalation Indicated voltage is < 105 VDC on vital 125 VDC buses 11DA 1 2 3 (Notes 1, 2) and 11DB for 15 min. - DIV II DG (DG 12)

Loss of (Note 1) SS2.1 None None Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, irradiated UNPLANNED loss of water level above irradiated fuel Vital DC Indicated voltage is < 105 VDC on vital 125 VDC buses 11DA Abnorm. fuel racks for 60 minutes or longer fuel Power and 11DB for 15 min. (Note 1)

Rad 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF

` UNPLANNED loss of Control Room indications for 15 minutes or UNPLANNED loss of Control Room indications for 15 minutes or Levels AG2.1 AS2.1 AA2.1 AU2.1 longer with a significant transient in progress longer

/ Rad Spent fuel pool level cannot be restored to at least 183 ft. Lowering of spent fuel pool level to 183 ft. (Level 3) IMMINENT uncovery of irradiated fuel in the REFUELING UNPLANNED water level drop in the REFUELING PATHWAY Effluent (Level 3) on G41R040A(B) for 60 min. (Note 1) on G41R040A(B) as indicated by Fuel Pool Drain Tank low water level alarm, 1 2 3 1 2 3 2

PATHWAY Table S-3 Significant Transients AA2.2 visual observation, water level drop in Upper Ctmt Pools, Aux SA3.1 SU3.1 Irradiated Release Point Table A-1 Effluent Monitor Classification Thresholds GE SAE Alert UE Damage to irradiated fuel resulting in a release of radioactivity AND Bldg Fuel Pools or the Fuel Transfer Canal AND UNPLANNED rise in corresponding area radiation levels as 3 None - Reactor scram

- UNPLANNED drop in reactor thermal power > 25%

An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room Fuel Event VALID alarm on any Table A-2 radiation monitor Loss of CR for 15 min. (Note 1) for 15 min. (Note 1) indicated by any of the following radiation monitors:

AA2.3 - Electrical load rejection > 25% electrical load AND SBGT A/B 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec - Ctmt 209 Airlock (1D21K630) Indications

- Ctmt Fuel Hdlg Area (1D21K626) - ECCS injection Any significant transient is in progress, Table S-3 Lowering of spent fuel pool level to 193 ft. (Level 2)

CTMT Vent Gaseous 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec on G41R040A(B) - Aux Bldg Fuel Hdlg Area(1D21K622)

- Thermal power oscillations > 10% RCS activity greater than Technical Specification allowable limits Radwaste Building Vent 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec 1 2 3 Turbine Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec SU4.1 Fuel Handling (Aux Bldg) Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown Table A-2 Fuel Damage Radiation Monitors 4 None Table S-2 Safety System Parameters Refer to Table F-1 for potential escalation due to degraded fuel clad barrier Offgas Pretreatment radiation monitor high-high alarm SU4.2 RCS - Reactor power Liquid Radwaste -------- -------- -------- 7.33E+5 cpm AA3.1 1 2 3 4 5 DEF Activity Coolant activity > 0.2 µCi/gm dose equivalent I-131 for > 48 3

- RPV water level Dose rate > 15 mR/hr in EITHER of the following areas:

• Ctmt Vent (P601-19A-G9) hours

- Control Room (SD21-K600) - RPV pressure OR

• FH Area Vent (P601-19A-C11)

- Central Alarm Station (by survey) - Containment pressure Coolant activity > 4.0 µCi/gm dose equivalent I-131 Table A-3 Safe Operation & Shutdown Rooms/Areas None

• Ctmt 209 Airlock (P844-1A-A1)

Area Rad instantaneous Levels Room / Area Mode - Suppression Pool water level AA3.2 3

• Ctmt Fuel Hdlg Area (P844-1A-A3)

- Suppression Pool temperature RCS leakage for 15 minutes or longer Control Building 111 SWGR Rms (0C202, 0C215) 3 An UNPLANNED event results in radiation levels that prohibit

• Aux Bldg Fuel Hdlg Area (P844-1A-A4) or IMPEDE access to any Table A-3 room or area (Note 5) 1 2 3 Auxiliary Building 93 RHR A Pump Room (1A103) 3 SU5.1 Auxiliary Building 93 RHR B Pump Room (1A105)

Auxiliary Building 93 Corridor (1A101) 3 3 HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or airborne attack threat within 30 minutes Confirmed SECURITY CONDITION or threat S 5 None None Refer to Table F-1 for potential escalation due to degraded RCS barrier RCS unidentified or pressure boundary leakage

> 10 gpm for 15 min. (Note 1)

RCS OR Auxiliary Building 119 Corridor (1A201) 3 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF RCS identified leakage > 25 gpm for 15 min. (Note 1)

System Leakage Auxiliary Building 119 RHR A Pump Room (1A203) 3 HS1.1 HA1.1 HU1.1 OR 1 Auxiliary Building 119 RHR B Pump Room (1A205)

None Auxiliary Building 119 RCIC Room (1A204) 3 3

A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by GGNS Security Shift A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROLLED AREA as reported by A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by GGNS Security Shift Supervision Malfunct. Leakage from the RCS to a location outside containment

> 25 gpm for 15 min. (Note 1)

Supervision GGNS Security Shift Supervision OR Security OR Notification of a credible security threat directed at the site Inability to shut down the reactor causing a challenge to RPV water Automatic or manual scram fails to shut down the reactor and Automatic or manual scram fails to shut down the reactor Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 A validated notification from NRC of an aircraft attack threat OR level or RCS heat removal subsequent manual actions taken at the reactor control consoles are Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 within 30 min. of the site A validated notification from the NRC providing information not successful in shutting down the reactor Radwaste Building 118 Radwaste Control Room (0R241) 3 of an aircraft threat 1 2 1 2 1 2 SS6.1 SA6.1 SU6.1 Seismic event greater than OBE levels An automatic or manual scram fails to shut down the reactor An automatic or manual scram fails to shut down the reactor An automatic scram did not shut down the reactor as 1 2 3 4 5 DEF as indicated by reactor power > 5% as indicated by reactor power > 5% indicated by reactor power > 5% after any RPS setpoint 2 HU2.1 AND AND is exceeded All actions to shut down the reactor are not successful as Manual scram actions taken at the reactor control console AND Refer to SA8.1 for potential escalation due to a seismic event Seismic event > OBE as indicated by annunciation of 6

indicated by reactor power > 5% (Mode Switch, Manual PBs, ARI/RPT) are not successful in A subsequent automatic scram or manual scram action None None EITHER of the following on SH13P856:

Seismic AND EITHER: shutting down the reactor as indicated by reactor power > 5% taken at the reactor control console (Mode Switch,

- Containment Operating Basis Earthquake (P856-1A-A3) None Event - RPV water level cannot be restored and maintained (Note 8) Manual PBs, ARI/RPT) is successful in shutting down the

- Drywell Operating Basis Earthquake (P856-1A-A5)

> -191 in. reactor as indicated by reactor power 5% (APRM RPS - Heat Capacity Temperature Limit (HCTL) exceeded downscale) (Note 8)

Failure (EP Figure 1)

Hazardous Event GGNS SU6.2 1 2 3 4 5 DEF Document Control A manual scram did not shut down the reactor as indicated by reactor power > 5% after any manual scram action (Mode NOTES HU3.1 A tornado strike within the PROTECTED AREA CONTROLLED COPY Switch or Manual PBs) was initiated Note 1: The Emergency Director should declare the event promptly upon AND determining that the time limit has been exceeded, or will likely be exceeded. The HU3.2 A subsequent automatic scram or manual scram action taken 3

Emergency Director is not allowed an additional 15 minutes to declare after the Internal room or area FLOODING of a magnitude sufficient at the reactor control console (Mode Switch, Manual PBs, time limit is exceeded. to require manual or automatic electrical isolation of a Table S-4 Communication Methods ARI/RPT) is successful in shutting down the reactor as SAFETY SYSTEM component required by Technical indicated by reactor power 5% (APRM downscale) (Note 8)

Note 2: If an ongoing release is detected and the release start time is unknown, State/

assume that the release duration has exceeded the specified time limit. Specifications for the current operating mode System Onsite NRC Natural or Refer to SA8.1 for potential escalation due to a hazardous event Local None Loss of all onsite or offsite communications capabilities Tech. HU3.3 Station Radio System X Note 3: If the effluent flow past an effluent monitor is known to have stopped due Hazard to actions to isolate the release path, then the effluent monitor reading is no longer Movement of personnel within the PROTECTED AREA is 1 2 3 VALID for classification purposes. IMPEDED due to an event external to the PROTECTED GGNS Plant Phone System X SU7.1 7

None AREA involving hazardous materials (e.g., an offsite chemical Public Address System X Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, Loss of all Table S-4 onsite communication methods spill or toxic gas release)

AS1.1 and AG1.1 should be used for emergency classification assessments until Emergency Notification System (ENS) X OR the results from a dose assessment using actual meteorology are available.

HU3.4 Loss of Commercial Telephone System X X Loss of all Table S-4 State and local agency communication Note 5: If the equipment in the listed room or area was already inoperable or out- A hazardous event that results in on-site conditions sufficient Comm. methods of-service before the event occurred, then no emergency classification is to prohibit the plant staff from accessing the site via personal Table F-2 EP-4 Aux Building Area Parameters Satellite Phones X X OR warranted. vehicles (Note 7) Operational Hotline X Loss of all Table S-4 NRC communication methods Area Operating Limit MAX SAFE Value Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30- FIRE potentially degrading the level of safety of the plant Area Temperature minute time limit, declaration of a General Emergency is not required.

MSL Pipe Tunnel Temp 185°F (P601-19A/18A-A3/A4) 250°F (E31-N604A, B, C, D, E, F) Hazardous event affecting SAFETY SYSTEMS needed for the Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, 1 2 3 4 5 DEF RHR-A Equip Area Temp 165°F (P601-20A-B1) 225°F (E31-N608A, N610A) current operating mode Refer to SA8.1 for potential escalation due to a fire Refer to Category H for potential Unusual Event due to a hazardous event ice, or vehicle breakdowns or accidents. 1 2 3 HU4.1 RHR-B Equip Area Temp 165°F (P601-20A-B1) 225°F (E31-N608B, N610B)

Note 8: A manual scram action is any operator action, or set of actions, which A FIRE is not extinguished within 15 min. of any of the RCIC Equip Area Temp 185°F (P601-21A-G3) 212°F (E31-N602A/B) SA8.1 causes the control rods to be rapidly inserted into the core, and does not include following FIRE detection indications (Note 1): RWCU-Pump Room 1 Temp 170°F (P680-11A-A1) NA The occurrence of any Table S-5 hazardous event Table S-5 Hazardous Events 8

manually driving in control rods or implementation of boron injection strategies.

- Report from the field (i.e., visual observation) RWCU-Pump Room 2 Temp 170°F (P680-11A-A2) NA AND Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of - Receipt of multiple (more than 1) fire alarms or indications Event damage has caused indications of degraded Table H-1 Fire Areas - Seismic event (earthquake) service before the hazardous event occurred, then emergency classification is not - Field verification of a single fire alarm Area Radiation performance on one train of a SAFETY SYSTEM needed warranted.

- Unit 1 Containment AND Hazardous RHR Room A 102 mr/hr (P844-1A-D4) 8 x 104 mr/hr for the current operating mode - Internal or external FLOODING event The FIRE is located within any Table H-1 area None Event 102 mr/hr (P844-1A-D4) 8 x 104 mr/hr AND EITHER:

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no - Unit 1 Auxiliary Building RHR Room B - High winds or tornado strike H 4 indications of degraded performance to at least one train of a SAFETY SYSTEM, HU4.2 Affecting

• Event damage has caused indications of

- Unit 1 Turbine Building RHR Hx A Hatch 102 mr/hr (P844-1A-C4) 8 x 104 mr/hr then this emergency classification is not warranted. Safety degraded performance to the second train of the - FIRE

- Control Building Receipt of a single fire alarm (i.e., no other indications of a SAFETY SYSTEM needed for the current Systems RHR Hx B Hatch 102 mr/hr (P844-1A-C4) 8 x 104 mr/hr Note 11: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in FIRE) (Note 11) operating mode - EXPLOSION

- Diesel Generator Rooms 102 mr/hr (P844-1A-D4) 8 x 104 mr/hr Fire None the containment or drywell. AND RCIC Room

- SSW Pump & Valve Rooms

• Event damage has resulted in VISIBLE DAMAGE - Other events with similar hazard The fire alarm is indicating a FIRE within any Table H-1 area Main Steam Line Rad Monitor Set Point Log (P601-19A-D4) 8 x 104 mr/hr to the second train of the SAFETY SYSTEM Hazards characteristics as determined by the AND SGTS Fltr Trn 2.5 mr/hr (P844-1A-C5) 8 x 102 mr/hr needed for the current operating mode Shift Manager The existence of a FIRE is not verified (i.e., proved or (Notes 9, 10)

None disproved) within 30 min. of alarm receipt (Note 1)

Table H-2 Unit Safe Operation & Shutdown Rooms/Areas F

FG1.1 1 2 3 FS1.1 1 2 3 FA1.1 1 2 3 HU4.3 Room / Area Mode A FIRE within the PROTECTED AREA not extinguished Loss of any two barriers Loss or potential loss of any two barriers (Table F-1) Any loss or any potential loss of either Fuel Clad or RCS Control Building 111 SWGR Rms (0C202, 0C215) 3 within 60 min. of the initial report, alarm or indication (Note 1) AND barrier (Table F-1)

Fission Loss or potential loss of the third barrier (Table F-1)

None Auxiliary Building 93 RHR A Pump Room (1A103) 3 HU4.4 Product Auxiliary Building 93 RHR B Pump Room (1A105) 3 A FIRE within the PROTECTED AREA that requires Barriers Auxiliary Building 93 Corridor (1A101) 3 firefighting support by an offsite fire response agency to `

extinguish Auxiliary Building 119 Corridor (1A201) 3 Gaseous release IMPEDING access to equipment necessary for Auxiliary Building 119 RHR A Pump Room (1A203) 3 Table F-1 Fission Product Barrier Threshold Matrix normal plant operations, cooldown or shutdown Auxiliary Building 119 RHR B Pump Room (1A205) 3 3 5 None Auxiliary Building 119 RCIC Room (1A204)

Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 3

HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas None Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB)

Hazardous Gas Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 into any Table H-2 room or area AND Loss Potential Loss Loss Potential Loss Loss Potential Loss Radwaste Building 118 Radwaste Control Room (0R241) 3 Entry into the room or area is prohibited or IMPEDED (Note 5)

A. RPV Water FCB1 SAP entry is required FCB2 RPV water level cannot RCB1 RPV water level cannot be CNB1 SAP entry is required Level be restored and restored and maintained Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to None None Room maintained > -167 in (TAF) > -167 in. (TAF) alternate locations 1 2 3 4 5 or cannot be determined or cannot be determined 1 2 3 4 5 DEF HS6.1 HA6.1 6 None An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel AND An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel None B. RCS Leak Rate RCB2 UNISOLABLE break in any of the following:

- Main steam line RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:

CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:

- One or more EP-4 MAX SAFE area Control - RCIC steam line - One or more EP-4 area radiation levels (Table F-2)

Control of any of the following key safety functions is not Room - RWCU radiation Operating Limits - One or more EP-4 MAX SAFE area reestablished within 15 min. (Note 1):

Evacuation - Feedwater (Table F-2) temperature levels (Table F-2) None

- Reactivity (Modes 1 and 2 only) None None

- RPV water level - HPCS - One or more EP-4 area

- RCS heat removal temperature Operating Limits RCB3 Emergency Depressurization (Table F-2) is required Other conditions exist that in the judgment of the Emergency Other conditions exist that in the judgment of the Emergency Director Other conditions exist that in the judgment of the Emergency Director Other conditions exist that in the judgment of the Emergency Director Director warrant declaration of a GENERAL EMERGENCY warrant declaration of a SITE AREA EMERGENCY warrant declaration of an ALERT warrant declaration of a UE 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF C. CTMT RCB5 Drywell pressure > 1.23 psig CNB3 UNPLANNED rapid drop in CNB5 Containment pressure > 15 psig Conditions due to RCS leakage Containment pressure following HG7.1 HS7.1 HA7.1 HU7.1 Containment pressure rise CNB6 Drywell or containment hydrogen 7

None None None concentration > 4%

Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director indicate that events are in progress or CNB4 Containment pressure response not have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the consistent with LOCA conditions CNB7 Parameters cannot be restored and ED core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to maintained within the safe zone of the Judgment containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of HCTL curve (EP Figure 1) actual loss of physical control of the facility. Releases can be malicious acts, (1) toward site personnel or equipment that site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring FCB3 Containment radiation RCB6 Drywell radiation CNB8 Containment radiation D. CTMT Rad /

reasonably expected to exceed EPA Protective Action could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY (RITS-K648B or C) > 400 R/hr (RITS-K648A or D) (RITS-K648B or C)

RCS Activity Guideline exposure levels offsite for more than the access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs. >100R/hr > 7,000 R/hr None None None immediate site area Any releases are not expected to result in exposure levels exposure levels. FCB4 Primary coolant activity which exceed EPA Protective Action Guideline exposure > 300 µCi/gm dose equivalent levels beyond the SITE BOUNDARY. I-131 Damage to a loaded cask CONFINEMENT BOUNDARY CNB9 UNISOLABLE direct downstream E. CTMT pathway to the environment exists after 1 2 3 4 5 DEF Integrity or Containment isolation signal Bypass None None None None None EU1.1 CNB10 Intentional Containment venting per Damage to a loaded cask CONFINEMENT BOUNDARY as EPs E

indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack) > EITHER of F. Emergency FCB5 Any condition in the opinion of FCB6 Any condition in the opinion RCB7 Any condition in the opinion of RCB8 Any condition in the opinion of CNB11 Any condition in the opinion of the CNB12 Any condition in the opinion of the None None None the following:: Director the Emergency Director that of the Emergency Director the Emergency Director that the Emergency Director that Emergency Director that indicates loss Emergency Director that ISFSI - 60 mrem/hr (gamma + neutron) on the top of the Judgment indicates loss of the Fuel Clad that indicates potential loss indicates loss of the RCS indicates potential loss of the of the Containment barrier indicates potential loss of the overpack barrier of the Fuel Clad barrier barrier RCS barrier Containment barrier

- 600 mrem/hr (gamma + neutron) on the side of the overpack (excluding inlet and outlet ducts)

Modes: I Power Operation 1

I I 2 Startup I I 3 I I 4 I I 5 I IDEF I Defueled

~

~ Entergy.

GGNS EAL- HOT MODES 1, 2, & 3 10-S-01-1 Date: 2/28/20 EPP 01-02 (Flowchart)

Page 1 of 2 Final (4/5/19)

Hot Shutdown Cold Shutdown Refueling

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment I EMERGENCY CLASSIFICATIONS EPP 01-02 (Flowchart)

See Flowchart EPP 01-02 Page 2 of 2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite dose Release of gaseous or liquid radioactivity greater than 2 times the Loss of RPV inventory affecting fuel clad integrity with containment Loss of RPV inventory affecting core decay heat removal Significant loss of RPV inventory UNPLANNED loss of RPV inventory than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE greater than 10 mrem TEDE or 50 mrem thyroid CDE ODCM limits for 60 minutes or longer challenged capability 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 4 5 4 5 4 5 4 5 AG1.1 AS1.1 AA1.1 AU1.1 CG1.1 CS1.1 CA1.1 CU1.1 Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor Reading on any Table A-1 effluent radiation monitor RPV level < -167 in. (TAF) for 30 min. (Note 1)

> column GE for 15 min. (Notes 1, 2, 3, 4) > column SAE for 15 min. (Notes 1, 2, 3, 4) > column ALERT for 15 min. (Notes 1, 2, 3, 4) > column UE for 60 min. (Notes 1, 2, 3) AND CONTAINMENT CLOSURE not established Loss of RPV inventory as indicated by RPV water level UNPLANNED loss of reactor coolant results in RPV water AND < -42 in. (Level 2) level less than a required lower limit for 15 min. (Note 1)

AA1.2 Any Containment Challenge indication, Table C-2 AG1.2 AS1.2 AU1.2 RPV water level < -150 in. (Level 1)

Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Sample analysis for a gaseous or liquid release indicates a CU1.2 CG1.2 CA1.2

> 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the concentration or release rate > 2 x ODCM limits for 60 min. RPV water level cannot be monitored 1

the SITE BOUNDARY (Note 4) the SITE BOUNDARY (Note 4) SITE BOUNDARY (Note 4) CS1.2 RPV water level cannot be monitored for 15 min. (Note 1)

(Notes 1, 2) RPV water level cannot be monitored for 30 min. (Note 1) AND EITHER 1 AG1.3 AS1.3 AA1.3 Analysis of a liquid effluent sample indicates a concentration RPV AND Core uncovery is indicated by any of the following:

CONTAINMENT CLOSURE established AND RPV water level < -167 in. (TAF)

AND EITHER

• UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory

• UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory

• Visual observation of UNISOLABLE RCS leakage Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or - UNPLANNED rise in Suppression Pool level of sufficient Rad beyond the SITE BOUNDARY: beyond the SITE BOUNDARY: or release rate that would result in doses > 10 mrem TEDE or Level magnitude to indicate core uncovery

• Visual observation of UNISOLABLE RCS leakage Effluent - Closed window dose rates > 1000 mR/hr expected to - Closed window dose rates > 100 mR/hr expected to 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for - Visual observation of UNISOLABLE RCS leakage of 60 min. of exposure (Notes 1, 2) CS1.3 continue for 60 min. continue for 60 min. sufficient magnitude to indicate core uncovery

- Analyses of field survey samples indicate thyroid CDE - Analyses of field survey samples indicate thyroid CDE - Containment/Drywell High Range Area Radiation Monitor RPV water level cannot be monitored for 30 min. (Note 1)

> 5000 mrem for 60 min. of inhalation > 500 mrem for 60 min. of inhalation AA1.4 (1D21-K648B-C) 100 R/hr AND (Notes 1, 2) (Notes 1, 2) AND Core uncovery is indicated by any of the following:

Field survey results indicate EITHER of the following at or Any Containment Challenge indication, Table C-2 - UNPLANNED rise in Suppression Pool level of sufficient beyond the SITE BOUNDARY:

magnitude to indicate core uncovery

- Closed window dose rates > 10 mR/hr expected to

- Visual observation of UNISOLABLE RCS leakage of continue for 60 min.

A sufficient magnitude to indicate core uncovery

- Analyses of field survey samples indicate thyroid CDE - Containment/Drywell High Range Area Radiation Monitor

> 50 mrem for 60 min. of inhalation (1D21-K648B-C) 100 R/hr (Notes 1, 2)

Loss of all offsite and all onsite AC power to ESF buses for Loss of all but one AC power source to ESF buses for 15 minutes Abnorm. Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, irradiated UNPLANNED loss of water level above irradiated fuel or longer 15 minutes or longer fuel racks for 60 minutes or longer fuel Rad 4 5 DEF 4 5 DEF 2

Levels 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF AG2.1 AS2.1 AU2.1 CA2.1 CU2.1

/ Rad AA2.1 Spent fuel pool level cannot be restored to at least 183 ft. Lowering of spent fuel pool level to 183 ft. (Level 3) UNPLANNED water level drop in the REFUELING PATHWAY None None Loss of all offsite and all onsite AC power to DIV I and DIV II AC power capability, Table C-3, to DIV I and DIV II ESF 4.16 C

Effluent IMMINENT uncovery of irradiated fuel in the REFUELING 2 (Level 3) on G41R040A(B) for 60 min. (Note 1) on G41R040A(B) PATHWAY as indicated by Fuel Pool Drain Tank low water level alarm, Loss of ESF 4.16 KV buses for 15 min. (Note 1) KV buses reduced to a single power source for 15 min.

visual observation, water level drop in Upper Ctmt Pools, Aux (Note 1)

ESF AC AND AA2.2 Bldg Fuel Pools or the Fuel Transfer Canal Table A-1 Effluent Monitor Classification Thresholds Power Any additional single power source failure will result in loss of Irradiated Damage to irradiated fuel resulting in a release of radioactivity AND UNPLANNED rise in corresponding area radiation levels as all AC power to SAFETY SYSTEMS Fuel Event Release Point GE SAE Alert UE AND Cold SD/

VALID alarm on any Table A-2 radiation monitor indicated by any of the following radiation monitors:

SBGT A/B 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec - Ctmt 209 Airlock (1D21K630) Refuel Inability to maintain plant in cold shutdown UNPLANNED rise in RCS temperature AA2.3

- Ctmt Fuel Hdlg Area (1D21K626) System CTMT Vent - Aux Bldg Fuel Hdlg Area(1D21K622) Malfunct. 4 5 4 5 Gaseous 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec Lowering of spent fuel pool level to 193 ft. (Level 2) on G41R040A(B) CA3.1 CU3.1 3

Radwaste Building Vent 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec UNPLANNED rise in RCS temperature to > 200°F for UNPLANNED rise in RCS temperature to > 200°F Turbine Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Radiation levels that IMPEDE access to equipment necessary for > Table C-4 duration (Note 1) normal plant operations, cooldown or shutdown None None OR CU3.2 Fuel Handling (Aux Bldg) Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Table A-2 Fuel Damage Radiation Monitors RCS UNPLANNED RPV pressure rise > 10 psig Temp. Loss of all RCS temperature and RPV water level AA3.1 1 2 3 4 5 DEF 3

indication for 15 min. (Note 1)

Liquid Radwaste -------- -------- -------- 7.33E+5 cpm Dose rate > 15 mR/hr in EITHER of the following areas:

• Ctmt Vent (P601-19A-G9)

- Control Room (SD21-K600)

• FH Area Vent (P601-19A-C11)

- Central Alarm Station (by survey)

Area Rad

• Ctmt 209 Airlock (P844-1A-A1)

Table A-3 Safe Operation & Shutdown Rooms/Areas None Levels 3 AA3.2

• Ctmt Fuel Hdlg Area (P844-1A-A3) Loss of vital DC power for 15 minutes or longer Room / Area Control Building 111 SWGR Rms (0C202, 0C215)

Mode 3

An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5)

• Aux Bldg Fuel Hdlg Area (P844-1A-A4) 4 None None None CU4.1 4 5 Auxiliary Building 93 RHR A Pump Room (1A103) 3 Loss of Indicated voltage is < 105 VDC on required vital 125 VDC Vital DC Auxiliary Building 93 RHR B Pump Room (1A105) 3 HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the SECURITY OWNER CONTROLLED Confirmed SECURITY CONDITION or threat buses 11DA and/or 11DB for 15 min. (Note 1)

Power Auxiliary Building 93 Corridor (1A101) AREA or airborne attack threat within 30 minutes 3

1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF Loss of all onsite or offsite communications capabilities Auxiliary Building 119 Corridor (1A201) 3 HS1.1 HA1.1 HU1.1 4 5 DEF 1 Auxiliary Building 119 RHR A Pump Room (1A203)

Auxiliary Building 119 RHR B Pump Room (1A205) 3 3

A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by GGNS Security Shift Supervision A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROLLED AREA as reported by GGNS Security Shift Supervision A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by GGNS Security Shift Supervision OR 5 None None None CU5.1 Loss of all Table C-5 onsite communication methods Security Auxiliary Building 119 RCIC Room (1A204) 3 OR Notification of a credible security threat directed at the site OR A validated notification from NRC of an aircraft attack threat OR Loss of Loss of all Table C-5 State and local agency communication Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 within 30 min. of the site A validated notification from the NRC providing information Comm. methods Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 of an aircraft threat OR Radwaste Building 118 Radwaste Control Room (0R241) 3 Loss of all Table C-5 NRC communication methods Seismic event greater than OBE levels 1 2 3 4 5 DEF Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode 2 None None Refer to CA6.1 for potential escalation due to a seismic event HU2.1 Seismic event > OBE as indicated by annunciation of EITHER of the following on SH13P856: CA6.1 4 5 6

Seismic - Containment Operating Basis Earthquake (P856-1A-A3) The occurrence of any Table C-6 hazardous event Event - Drywell Operating Basis Earthquake (P856-1A-A5) AND Event damage has caused indications of degraded Hazardous performance on one train of a SAFETY SYSTEM needed None None for the current operating mode Refer to Category H for potential Unusual Event due to a hazardous event Hazardous Event Event Affecting AND EITHER:

1 2 3 4 5 DEF

• Event damage has caused indications of Safety NOTES HU3.1 A tornado strike within the PROTECTED AREA Systems degraded performance to the second train of the SAFETY SYSTEM needed for the current Note 1: The Emergency Director should declare the event promptly upon operating mode determining that the time limit has been exceeded, or will likely be exceeded. The HU3.2

• Event damage has resulted in VISIBLE DAMAGE Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Internal room or area FLOODING of a magnitude sufficient to the second train of the SAFETY SYSTEM 3

to require manual or automatic electrical isolation of a needed for the current operating mode Note 2: If an ongoing release is detected and the release start time is unknown, SAFETY SYSTEM component required by Technical (Notes 9, 10) assume that the release duration has exceeded the specified time limit. Specifications for the current operating mode None Refer to CA6.1 for potential escalation due to a hazardous event Natural or Note 3: If the effluent flow past an effluent monitor is known to have stopped due HU3.3 Tech. to actions to isolate the release path, then the effluent monitor reading is no longer Movement of personnel within the PROTECTED AREA is VALID for classification purposes. Table C-1 Sumps / Pool Table C-3 AC Power Sources Table C-5 Communication Methods Table C-6 Hazardous Events Hazard IMPEDED due to an event external to the PROTECTED None Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AREA involving hazardous materials (e.g., an offsite chemical State/

AS1.1 and AG1.1 should be used for emergency classification assessments until spill or toxic gas release) - Drywell equipment drain sump System Onsite NRC - Seismic event (earthquake)

Offsite Local the results from a dose assessment using actual meteorology are available. - Internal or external FLOODING event HU3.4 - Drywell floor drain sump - ESF Transformer 11 Station Radio System X Note 5: If the equipment in the listed room or area was already inoperable or out- A hazardous event that results in on-site conditions sufficient - CTMT equipment drain sump - High winds or tornado strike of-service before the event occurred, then no emergency classification is - ESF Transformer 12 GGNS Plant Phone System X to prohibit the plant staff from accessing the site via personal warranted. vehicles (Note 7) - CTMT floor drain sump - ESF Transformer 21 Public Address System X - FIRE Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30- - Suppression Pool Emergency Notification System (ENS) X - EXPLOSION FIRE potentially degrading the level of safety of the plant Onsite minute time limit, declaration of a General Emergency is not required. - RHR A, B, C, HPCS, LPCS, RCIC room sumps

- DIV I DG (DG 11) Commercial Telephone System X X - Other events with similar hazard Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, 1 2 3 4 5 DEF - Auxiliary Building floor drain sump characteristics as determined by the ice, or vehicle breakdowns or accidents. - DIV II DG (DG 12) Satellite Phones X X Shift Manager Refer to CA6.1 for potential escalation due to a fire HU4.1 Operational Hotline X Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include A FIRE is not extinguished within 15 min. of any of the manually driving in control rods or implementation of boron injection strategies. following FIRE detection indications (Note 1):

- Report from the field (i.e., visual observation)

Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of - Receipt of multiple (more than 1) fire alarms or indications Table C-2 Containment Challenge Indications service before the hazardous event occurred, then emergency classification is not Table H-1 Fire Areas Table C-4 RCS Heat-up Duration Thresholds

- Field verification of a single fire alarm warranted.

- Unit 1 Containment AND The FIRE is located within any Table H-1 area - CONTAINMENT CLOSURE not established (Note 6) RCS Status CONTAINMENT Heat-up Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no - Unit 1 Auxiliary Building CLOSURE Status Duration H

indications of degraded performance to at least one train of a SAFETY SYSTEM, 4 then this emergency classification is not warranted. - Unit 1 Turbine Building HU4.2 - Drywell or containment hydrogen concentration > 4%

Receipt of a single fire alarm (i.e., no other indications of a Intact N/A 60 min.*

- Control Building .

Note 11: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in FIRE) (Note 11) - UNPLANNED rise in containment pressure None the containment or drywell. - Diesel Generator Rooms 20 min.*

Fire AND established

- SSW Pump & Valve Rooms The fire alarm is indicating a FIRE within any Table H-1 area - Exceeding one or more Auxiliary Building Control Not intact Hazards AND MAX SAFE area radiation levels (EP-4) not established 0 min.

The existence of a FIRE is not verified (i.e., proved or None disproved) within 30 min. of alarm receipt (Note 1)

• If an RCS heat removal system is in operation within this time frame and Table H-2 Unit Safe Operation & Shutdown Rooms/Areas RCS temperature is being reduced, the EAL is not applicable HU4.3 Room / Area Mode A FIRE within the PROTECTED AREA not extinguished within Control Building 111 SWGR Rms (0C202, 0C215) 3 60 min. of the initial report, alarm or indication (Note 1)

Auxiliary Building 93 RHR A Pump Room (1A103) 3 HU4.4 Auxiliary Building 93 RHR B Pump Room (1A105) 3 A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to Auxiliary Building 93 Corridor (1A101) 3 extinguish Auxiliary Building 119 Corridor (1A201) 3 3 Gaseous release IMPEDING access to equipment necessary for Auxiliary Building 119 RHR A Pump Room (1A203) normal plant operations, cooldown or shutdown Auxiliary Building 119 RHR B Pump Room (1A205) 3 3

5 None Auxiliary Building 119 RCIC Room (1A204)

Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 3

HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas None Hazardous Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 into any Table H-2 room or area Gas AND Radwaste Building 118 Radwaste Control Room (0R241) 3 Entry into the room or area is prohibited or IMPEDED (Note 5)

Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to Room alternate locations 1 2 3 4 5 1 2 3 4 5 DEF HS6.1 GGNS HA6.1 6 None An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel AND An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel None Document Control CONTROLLED COPY Control Control of any of the following key safety functions is not Room reestablished within 15 min. (Note 1): #____________

Evacuation

- Reactivity (Modes 1 and 2 only)

- RPV water level

- RCS heat removal Other conditions exist that in the judgment of the Emergency Other conditions exist that in the judgment of the Emergency Director Other conditions exist that in the judgment of the Emergency Director Other conditions exist that in the judgment of the Emergency Director Director warrant declaration of a GENERAL EMERGENCY warrant declaration of a SITE AREA EMERGENCY warrant declaration of an ALERT warrant declaration of a UE 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF 1 2 3 4 5 DEF HG7.1 HS7.1 HA7.1 HU7.1 7 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the ED core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to Judgment containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of actual loss of physical control of the facility. Releases can be malicious acts, (1) toward site personnel or equipment that site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring reasonably expected to exceed EPA Protective Action could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Guideline exposure levels offsite for more than the access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs.

immediate site area Any releases are not expected to result in exposure levels exposure levels.

which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

Damage to a loaded cask CONFINEMENT BOUNDARY 1 2 3 4 5 DEF EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as E

indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack) > EITHER None None None of the following:

ISFSI - 60 mrem/hr (gamma + neutron) on the top of the overpack

- 600 mrem/hr (gamma + neutron) on the side of the overpack (excluding inlet and outlet ducts)

Modes: D 1 Power Operation D

Startup 2

D 3 D 4 D 5 DDEF Defueled

... ~ Entergy_ GGNS EAL - COLD MODES 4, 5 & Defueled 10-S-01-1 Date: 2/28/20 EPP 01-02 (Flowchart)

Page 2 of 2 Final (4/5/19)

Hot Shutdown Cold Shutdown Refueling

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 1 of 245 EAL Technical Bases TABLE OF CONTENTS

1.0 INTRODUCTION

..2 2.0 DISCUSSION.2 2.1 Background....2 2.2 Fission Product Barriers......3 2.3 Fission Product Barrier Classification Criteria..3 2.4 EAL Organization......4 2.5 Technical Bases Information......6 2.6 Operating Mode Applicability......7 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ..8 3.1 General Considerations.......8 3.2 Classification Methodology.....9

4.0 REFERENCES

................................................................................................................................................... 13 4.1 Developmental ...........13 4.2 Implementing.......13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ............................................................................ ..14 5.1 Definitions (ref. 4.1.1 except as noted) ..14 5.2 Abbreviations/Acronyms....28 6.0 GGNS-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE.............................................................................. ..21 Category A - Abnormal Rad Levels / Rad Effluent...26 Category C - Cold Shutdown / Refueling System Malfunction...64 Category E - Independent Spent Fuel Storage Installation (ISFSI) ....106 Category F - Fission Product Barrier Degradation.109 Table F-1 Fission Product Barrier Threshold Matrix and Bases....116 Category H - Hazards and Other Conditions Affecting Plant Safety...168 Category S - System Malfunction.....205 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 2 of 245 EAL Technical Bases

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Grand Gulf Nuclear Station (GGNS). It should be used to facilitate review of the GGNS EALs and provide historical documentation for future reference.

Decision-makers responsible for implementation of 10-S-01-1, Activation of the Emergency Plan, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION

2.1 Background

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the GGNS Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 Methodology for Development of Emergency Action Levels as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).

Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 3 of 245 EAL Technical Bases Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ref. 4.1.1), GGNS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based.

That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies a greater probability of barrier loss and reduced certainty of maintaining the barrier.

The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves.

C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency.

2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

Alert:

Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 4 of 245 EAL Technical Bases 2.4 EAL Organization The GGNS EAL scheme includes the following features:

Division of the EAL set into three broad groups:

o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode.

o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.

The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition.

This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The GGNS EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the GGNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The GGNS EAL categories and subcategories are listed below.

The primary tool for determining the emergency classification level is the EAL Classification Matrix.

The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 5 of 245 EAL Technical Bases EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I

Any Operating Mode:

A - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)

Hot Conditions:

S - System Malfunction 1 - Loss of ESF AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System 1 - RPV Level Malfunction 2 - Loss of ESF AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 6 of 245 EAL Technical Bases 2.5 Technical Bases Information EAL technical bases are provided in Attachment 2 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

Category Letter & Title Subcategory Number & Title Initiating Condition (IC)

Site-specific description of the generic IC given in NEI 99-01 Rev. 6.

EAL Identifier (enclosed in rectangle)

Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter): Corresponds to the EAL category as described above (A, C, E, F, H or S)

2. Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 7 of 245 EAL Technical Bases Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 -

Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions)

Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section.

These definitions can also be found in Section 5.1.

Basis:

An EAL basis section that provides GGNS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

Reference(s):

Source documentation from which the EAL is derived 2.6 Operating Mode Applicability 1 Power Operation Reactor is critical and the mode switch is in RUN 2 Startup The mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is >200ºF 4 Cold Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is 200ºF 5 Refueling The mode switch is in REFUEL or SHUTDOWN with one or more reactor vessel head closure bolts are less than fully tensioned DEF Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 8 of 245 EAL Technical Bases 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices.

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref.

4.1.8).

3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 9 of 245 EAL Technical Bases 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 4.1.4).

3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started.

When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8).

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 10 of 245 EAL Technical Bases 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met a Site Area Emergency should be declared.

There is no additive effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met an Alert should be declared.

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.

3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met.

While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

3.2.4 Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met.

3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 11 of 245 EAL Technical Bases or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.

3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.

In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 12 of 245 EAL Technical Bases In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 13 of 245 EAL Technical Bases

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.

4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 GGNS Technical Specifications Table 1.1-1, Modes 4.1.7 GGNS Offsite Dose Calculation Manual (ODCM) 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 GGNS Emergency Plan 4.1.10 GGNS UFSAR 9.1.4.2.10.4 Storage of Fuel at the Independent Spent Fuel Storage Installation (ISFSI) 4.1.11 GGNS UFSAR 9.1.4.2.10 Description of Fuel Transfer 4.1.12 SOPP-018-1 Shutdown Operations Protection Plan 4.1.13 10-S-01-12 Radiological Assessment and Protective Action Recommendations 4.2 Implementing 4.2.1 10-S-01-1 Activation of the Emergency Plan 4.2.2 NEI 99-01 Rev. 6 to GGNS EAL Comparison Matrix 4.2.3 GGNS EAL Matrix J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 14 of 245 EAL Technical Bases 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

Alert - Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Confinement Boundary - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC) (ref. 4.1.10).

Containment Closure - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established (ref. 4.1.12).

Emergency Action Level (EAL) - A pre-determined, site-specific, observable threshold for an INITIATING CONDITION that, when met or exceeded, places the plant in a given emergency classification level.

Emergency Classification Level (ECL) - One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)

Alert Site Area Emergency (SAE)

General Emergency (GE)

Explosion - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.)

should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 15 of 245 EAL Technical Bases Fire - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

Fission Product Barrier Threshold - A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

Flooding - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

General Emergency - Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Hostage - A person(s) held as leverage against the station to ensure that demands will be met by the station.

Hostile Action - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

Hostile Force - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

Imminent - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Impede(d) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Independent Spent Fuel Storage Installation (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Initiating Condition (IC) - An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 16 of 245 EAL Technical Bases Projectile - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

Protected Area - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled. (ref. 4.1.9).

RCS Intact - The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Refueling Pathway - Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway (ref. 4.1.11).

Restore - Take the appropriate action required to return the value of an identified parameter to the applicable limits.

Safety System - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a HOSTILE ACTION.

Security Owner Controlled Area (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary.

Site Area Emergency - Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 17 of 245 EAL Technical Bases Site Boundary - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor (ref. 4.1.13)

Unisolable - An open or breached system line that cannot be isolated, remotely or locally.

Unplanned - A parameter change or an event that is not 1) the result of an intended evolution or

2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Unusual Event - Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Valid - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 18 of 245 EAL Technical Bases 5.2 Abbreviations/Acronyms

°F ....................................................................................................... Degrees Fahrenheit

° ........................................................................................................................... Degrees AB ........................................................................................................... Auxiliary Building AC........................................................................................................ Alternating Current AOP ................................................................................. Abnormal Operating Procedure APRM .................................................................................. Average Power Range Meter ARI................................................................................................ Alternate Rod Insertion ATWS ..................................................................... Anticipated Transient Without Scram BWR ............................................................................................... Boiling Water Reactor BWROG .................................................................Boiling Water Reactor Owners Group CDE ...................................................................................... Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CNB ................................................................................................... Containment Barrier CS.................................................................................................................... Core Spray CTMT............................................................................................................ Containment DEF .................................................................................................................... Defueled DBA ............................................................................................... Design Basis Accident DC ...............................................................................................................Direct Current D/G ......................................................................................................... Diesel Generator EAL ............................................................................................. Emergency Action Level ECCS............................................................................ Emergency Core Cooling System ECL.................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP .............................................................................. Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPG ............................................................................... Emergency Procedure Guideline EPP ....................................................................................... Emergency Plan Procedure ERO .......................................................................... Emergency Response Organization ESF......................................................................................... Engineered Safety Feature FAA.................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FCB ........................................................................................................ Fuel Clad Barrier FEMA............................................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency HCTL ............................................................................ Heat Capacity Temperature Limit HPCS....................................................................................... High Pressure Core Spray IC ......................................................................................................... Initiating Condition IPEEE ................ Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI............................................................ Independent Spent Fuel Storage Installation Keff ........................................................................ Effective Neutron Multiplication Factor LCO ..................................................................................Limiting Condition of Operation LER................................................................................................Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPCS ........................................................................................ Low Pressure Core Spray J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 19 of 245 EAL Technical Bases LRW ....................................................................................................... Liquid Radwaste LWR .................................................................................................. Light Water Reactor MPC................................... Maximum Permissible Concentration/Multi-Purpose Canister MPH........................................................................................................... Miles Per Hour mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MSCRWL ...................................................... Minimum Steam Cooling RPV Water Level MSIV ...................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line MW .................................................................................................................... Megawatt NEI............................................................................................... Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NESP ................................................................... National Environmental Studies Project NORAD ................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System OBE ...................................................................................... Operating Basis Earthquake ODCM.............................................................................Offsite Dose Calculation Manual ONEP ...................................................................................Off-Normal Event Procedure ORO ................................................................................. Offsite Response Organization PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PB .................................................................................................................... Pushbutton PCIS ..................................................................... Primary Containment Isolation System PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PSIG ............................................................................... Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen RCB ................................................................................................................RCS Barrier RCIC ................................................................................. Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RETS .........................................................Radiological Effluent Technical Specifications RHR ............................................................................................. Residual Heat Removal RPS ........................................................................................ Reactor Protection System RPT .............................................................................................Recirculation Pump Trip RPV ........................................................................................... Reactor Pressure Vessel RWCU ......................................................................................... Reactor Water Cleanup SAP ....................................................................................... Severe Accident Procedure SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ...................................................................... Self-Contained Breathing Apparatus SOCA ..............................................................................Security Owner Controlled Area SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SRV .....................................................................................................Safety Relief Valve J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 20 of 245 EAL Technical Bases SSE ....................................................................................... Safe Shutdown Earthquake TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center UFSAR .................................................................. Updated Final Safety Analysis Report USGS ............................................................................ United States Geological Survey J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 21 of 245 EAL Technical Bases 6.0 GGNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a GGNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the GGNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

GGNS NEI 99-01 Rev. 6 Example EAL IC EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 AG2.1 AG2 1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 22 of 245 EAL Technical Bases GGNS NEI 99-01 Rev. 6 Example EAL IC EAL CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 23 of 245 EAL Technical Bases GGNS NEI 99-01 Rev. 6 Example EAL IC EAL HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 N/A SU7 1, 2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 24 of 245 EAL Technical Bases GGNS NEI 99-01 Rev. 6 Example EAL IC EAL SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 26 of 245 Attachment 2 Emergency Action Level Technical Bases Category A - Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 27 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

AU1.1 Unusual Event Reading on any Table A-1 effluent radiation monitor > column "UE" for 60 min.

(Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec CTMT Vent Gaseous 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec Radwaste Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent Liquid 7.33E+5 cpm Radwaste ---- ---- ----

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 28 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses a potential reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such releases are typically associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. IAS-04-1-01-D17-1 Process Radiation Monitoring J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 29 of 245 Attachment 2 Emergency Action Level Technical Bases

2. Offsite Dose Calculation Manual

3. XC-Q1D17-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values

4. NEI 99-01 AU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 30 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL:

AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

This IC addresses a potential reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 31 of 245 Attachment 2 Emergency Action Level Technical Bases Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC AA1.

Reference(s):

1. Offsite Dose Calculation Manual

2. NEI 99-01 AU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 32 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

AA1.1 Alert Reading on any Table A-1 effluent radiation monitor > column "ALERT" for 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Note 4 The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec CTMT Vent Gaseous 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec Radwaste Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent Liquid 7.33E+5 cpm Radwaste ---- ---- ----

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 33 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. IAS-04-1-01-D17-1 Process Radiation Monitoring

2. XC-Q1D17-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values

3. NEI 99-01 AA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 34 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AS1.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 35 of 245 Attachment 2 Emergency Action Level Technical Bases Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 36 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 37 of 245 Attachment 2 Emergency Action Level Technical Bases This EAL is assessed per the ODCM (ref. 2)

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. Offsite Dose Calculation Manual

3. NEI 99-01 AA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 38 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 39 of 245 Attachment 2 Emergency Action Level Technical Bases conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 40 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor > column "SAE" for 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec CTMT Vent Gaseous 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec Radwaste Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent Liquid 7.33E+5 cpm Radwaste ---- ---- ----

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 41 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. IAS-04-1-01-D17-1 Process Radiation Monitoring

2. XC-Q1D17-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values

3. NEI 99-01 AS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 42 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AG1.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 43 of 245 Attachment 2 Emergency Action Level Technical Bases Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 44 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

AS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 100 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 45 of 245 Attachment 2 Emergency Action Level Technical Bases The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC AG1.

Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 46 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor > column "GE" for 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point GE SAE Alert UE 8.1E+2 Ci/sec 8.1E+1 Ci/sec 8.1E+0 Ci/sec 6.7E-2 Ci/sec SBGT A/B 6.4E+2 Ci/sec 6.4E+1 Ci/sec 6.4E+0 Ci/sec 6.7E-2 Ci/sec CTMT Vent Gaseous 5.1E+1 Ci/sec 5.1E+0 Ci/sec 5.1E-1 Ci/sec 6.7E-2 Ci/sec Radwaste Building Vent 1.3E+1 Ci/sec 1.3E+0 Ci/sec 1.3E-1 Ci/sec 6.7E-2 Ci/sec Turbine Building Vent 8.6E+3 Ci/sec 8.6E+2 Ci/sec 8.6E+1 Ci/sec 6.7E-2 Ci/sec Fuel Handling (Aux BLDG) Vent Liquid 7.33E+5 cpm Radwaste ---- ---- ----

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 47 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Reference(s):

1. IAS-04-1-01-D17-1 Process Radiation Monitoring

2. XC-Q1D17-17001 Grand Gulf Nuclear Station (GGNS) Radiological Effluent EAL Threshold Values

3. NEI 99-01 AG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 48 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

AG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs AA1.1, AS1.1 and AG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 49 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

AG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 1,000 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 50 of 245 Attachment 2 Emergency Action Level Technical Bases The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Reference(s):

1. 10-S-01-12 Radiological Assessment and Protective Action Recommendations

2. NEI 99-01 AG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 51 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL:

AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by Fuel Pool Drain Tank low water level alarm, visual observation, water level drop in Upper Ctmt Pools, Aux Bldg Fuel Pools or the Fuel Transfer Canal AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

Ctmt 209 Airlock (1D21K630)

Ctmt Fuel Hdlg Area (1D21K626)

Aux Bldg Fuel Hdlg Area(1D21K622)

Mode Applicability:

All Definition(s):

UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

REFUELING PATHWAY- Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway.

Basis:

This IC addresses a drop in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level drop will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 52 of 245 Attachment 2 Emergency Action Level Technical Bases the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations.

The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AA2.

Reference(s):

1. 05-1-02-II-8 High Radiation During Fuel Handling

2. 04-1-01-D21-1 Area Radiation Monitoring System

3. UFSAR 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation

4. NEI 99-01 AU2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 53 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

All Definition(s):

CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC).

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

REFUELING PATHWAY- Reactor cavity (well), upper containment pool, fuel transfer canal, and auxiliary building fuel pools, but not including the reactor vessel, comprise the refueling pathway.

Basis:

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the REFUELING PATHWAY. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL escalates from AU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 54 of 245 Attachment 2 Emergency Action Level Technical Bases While an area radiation monitor could detect a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC AS1 .

Reference(s):

1. NEI 99-01 AA2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 55 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND VALID alarm on any of the following radiation monitors:

Ctmt Vent (P601-19A-G9)

FH Area Vent (P601-19A-C11)

Ctmt 209 Airlock (P844-1A-A1)

Ctmt Fuel Hdlg Area (P844-1A-A3)

Aux Bldg Fuel Hdlg Area (P844-1A-A4)

Mode Applicability:

All Definition(s):

CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC).

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This EAL addresses events that have caused actual damage to an irradiated fuel assembly.

These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 56 of 245 Attachment 2 Emergency Action Level Technical Bases This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EU1.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be via IC AS1.

Reference(s):

1. 05-1-02-II-8 High Radiation During Fuel Handling

2. 04-1-01-D21-1 Area Radiation Monitoring System

3. UFSAR 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation

4. Offsite Dose Calculation Manual

5. NEI 99-01 AA2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 57 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

AA2.3 Alert Lowering of spent fuel pool level to 193 ft. (Level 2) on G41R040A(B)

Mode Applicability:

All Definition(s):

None Basis:

This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via IC AS1 or AS2.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 193 ft. 2 1/8 in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 183 ft. 2 1/8 in. rounded to 183 ft. for readability) (ref. 1).

G41R040A9B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148 Control Building in the Lower Cable Spreading Room.

Reference(s):

1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling

2. NEI 99-01 AA2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 58 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

AS2.1 Site Area Emergency Lowering of spent fuel pool level to 183 ft. (Level 3) on G41R040A(B)

Mode Applicability:

All Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 193 ft. 2 1/8 in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 183 ft. 2 1/8 in. rounded to 183 ft. for readability) (ref. 1).

G41R040A9B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148 Control Building in the Lower Cable Spreading Room.

Reference(s):

1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling

2. NEI 99-01 AS2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 59 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 183 ft. (Level 3) on G41R040A(B) for 60 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All Definition(s):

None Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 193 ft. 2 1/8 in. rounded to 193 ft. for readability) and SFP level at the top of the fuel racks (Level 3 - 183 ft. 2 1/8 in. rounded to 183 ft. for readability) (ref. 1).

G41R040A9B) Spent Fuel Pool Level Instrument is not located in the Control Room. The display cabinets are located in the 148 Control Building in the Lower Cable Spreading Room.

Reference(s):

1. 05-S-01-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling

2. NEI 99-01 AG2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 60 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas:

Control Room (SD21-K600)

Central Alarm Station (by survey)

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis:

Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room is monitored for excessive radiation by one detector, SD21-K600 (ref.

1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. While the CAS is in the Control Room Envelope, there are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area.

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

Reference(s):

1. 06-IC-1D21-R-1001 Area Radiation Monitoring Calibration

2. NEI 99-01 AA3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 61 of 245 Attachment 2 Emergency Action Level Technical Bases Category: A - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5)

Note 5: If the necessary equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table A-3 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Control Building 111 SWGR Rms (0C202, 0C215) 3 Auxiliary Building 93 RHR A Pump Room (1A103) 3 Auxiliary Building 93 RHR B Pump Room (1A105) 3 Auxiliary Building 93 Corridor (1A101) 3 Auxiliary Building 119 Corridor (1A201) 3 Auxiliary Building 119 RHR A Pump Room (1A203) 3 Auxiliary Building 119 RHR B Pump Room (1A205) 3 Auxiliary Building 119 RCIC Room (1A204) 3 Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 Radwaste Building 118 Radwaste Control Room (0R241) 3 Mode Applicability:

3 - Hot Shutdown Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 62 of 245 Attachment 2 Emergency Action Level Technical Bases UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the rise in radiation levels and determine if another IC may be applicable.

For AA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the higher radiation levels. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

An emergency declaration is not warranted if any of the following conditions apply:

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.

The higher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g.,

radiography, spent filter or resin transfer, etc.).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 63 of 245 Attachment 2 Emergency Action Level Technical Bases measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3 (Mode 3 only).

Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases

2. NEI 99-01 AA3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 64 of 245 Attachment 2 Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200ºF); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refueling, DEF - Defueled).

The events of this category pertain to the following subcategories:

1. RPV Level RPV water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV ESF buses.

3. RCS Temperature Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 65 of 245 Attachment 2 Emergency Action Level Technical Bases

5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 66 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:

CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Grand Gulf is equipped with multiple RPV water level instruments including: Wide Range, Fuel Zone, Shutdown Range, Upset Range, and Narrow Range (ref. 1). Multiple instruments on different reference and variable legs should be monitored. The Upset Range and Shutdown Range instruments share a common reference leg; therefore, Narrow Range instruments should be routinely monitored when relying on Shutdown or Upset Range instrument as the primary indication.

With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 11.4 in. (ref. 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.

With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange. Technical Specifications require at least 22 ft 8 in. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations (ref. 3). The RPV flange is at approximately 212 in. on the Shutdown Range. (ref. 4).

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 67 of 245 Attachment 2 Emergency Action Level Technical Bases This EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s):

1. 02-S-01-40 EP Technical Bases

2. 05-S-01-EP-2 RPV Control

3. Technical Specifications 3.9.6

4. 07-S-14-413 RPV Disassembly

5. NEI 99-01 CU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 68 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:

CU1.2 Unusual Event RPV water level cannot be monitored AND EITHER UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps/Pool Drywell equipment drain sump Drywell floor drain sump CTMT equipment drain sump CTMT floor drain sump Suppression Pool RHR A, B, C, HPCS, LPCS, RCIC room sumps Auxiliary Building floor drain sump Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

UNPLANNED-. A parameter changes or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 69 of 245 Attachment 2 Emergency Action Level Technical Bases In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1).

In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rise must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2, 3).

An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory.

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

This EAL addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Reference(s):

1. 03-1-01-5 Refueling

2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601

3. 04-1-02-1H13-P680 Alarm Response Instruction Panel 1H13-P680

4. 05-S-01-EP-4 Auxiliary Building Control

5. NEI 99-01 CU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 70 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant Loss of RPV inventory EAL:

CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < -42 in. (Level 2)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

None Basis:

The threshold RPV water level of -42 in. is the Level 2 actuation setpoint for HPCS and RCIC.

Although RCIC cannot restore RPV inventory in the cold condition, the Level 2 actuation setpoint is operationally significant and is indicative of a loss of RPV inventory significantly below the low RPV water level scram setpoint specified in CU1.1 (ref. 1, 2).

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.

Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

Reference(s):

1. Technical Specifications Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. 04-1-02-1H13-P601-16A-A4 Alarm Response Instruction Panel 1H13-P601 panel 16A-A4 RX LVL 2 (-42") LO

3. NEI 99-01 CA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 71 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant Loss of RPV inventory EAL:

CA1.2 Alert RPV water level cannot be monitored for 15 min. (Note 1)

AND EITHER UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table C-1 Sumps/Pool Drywell equipment drain sump Drywell floor drain sump CTMT equipment drain sump CTMT floor drain sump Suppression Pool RHR A, B, C, HPCS, LPCS, RCIC room sumps Auxiliary Building floor drain sump Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 72 of 245 Attachment 2 Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1).

In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2, 3).

An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory.

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

Reference(s):

1. 03-1-01-5 Refueling

2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601

3. 04-1-02-1H13-P680 Alarm Response Instruction Panel 1H13-P680 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 73 of 245 Attachment 2 Emergency Action Level Technical Bases

4. 05-S-01-EP-4 Auxiliary Building Control

5. NEI 99-01 CA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 74 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV water level < -150 in. (Level 1)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

The threshold RPV water level of -150 in. is the low-low-low ECCS actuation setpoint (Level 1).

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further lowering of RPV water level and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier (ref. 1, 2).

This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 75 of 245 Attachment 2 Emergency Action Level Technical Bases Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RPV levels of EALs CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. Technical Specifications Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. 04-1-02-1H13-P601-17A-E2 Alarm Response Instruction Panel 1H13-P601 panel 17A-E2 RX LVL 1 (-150") LO

3. NEI 99-01 CS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 76 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV water level < -167 in. (TAF)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of -167 in.), core uncovery starts to occur (ref. 1).

This IC addresses a significant and prolonged loss of RPV level control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 77 of 245 Attachment 2 Emergency Action Level Technical Bases The difference in the specified RPV levels of EALs CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1.

Reference(s):

1. 02-S-01-40 EP Technical Bases

2. NEI 99-01 CS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 78 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RPV level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED rise in Suppression Pool level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery Containment High Range Area Radiation Monitor (1D21-K648B-C) 100R/hr Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change, or event may be known or unknown.

Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1).

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 79 of 245 Attachment 2 Emergency Action Level Technical Bases In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in Suppression Pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory.

In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise, with corresponding indications on area radiation monitors. 100R/hr is used for this indication on Containment High Range Radiation Monitors (1D21-K648B and C. These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity elevation.

This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

In this EAL, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or AG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 80 of 245 Attachment 2 Emergency Action Level Technical Bases Reference(s):

1. 03-1-01-5 Refueling

2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601

3. 04-1-02-1H13-P680 Alarm Response Instruction Panel 1H13-P680

4. 05-S-01-EP-4 Auxiliary Building Control

5. 06-IC-1D21-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration

6. NEI 99-01 CS1

7. Calculation J-D21-1, Set Points Determination for High Range DW & Containment Radiation Monitors (D21 System)

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 81 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:

CG1.1 General Emergency RPV level < -167 in. (TAF) for 30 min. (Note 1)

AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

Drywell or containment hydrogen concentration > 4%

UNPLANNED rise in containment pressure Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 82 of 245 Attachment 2 Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

When RPV level drops below -167 in., core uncovery starts to occur (ref. 1).

Four conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE is not established.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4%

(ref. 2).

Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.

Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 3).

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 83 of 245 Attachment 2 Emergency Action Level Technical Bases established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Reference(s):

1. 02-S-01-40 EP Technical Bases

2. BWROG Emergency Procedure and Severe Accident Guidelines, Revision 3, p. B-16-64

3. 05-S-01-EP-4, Auxiliary Building Control

4. NEI 99-01 CG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 84 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:

CG1.2 General Emergency RPV level cannot be monitored for 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED rise in Suppression Pool level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery Containment/Drywell High Range Area Radiation Monitor (1D21-K648B-C) 100R/hr AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-2 Containment Challenge Indications CONTAINMENT CLOSURE not established (Note 6)

Drywell or containment hydrogen concentration > 4%

UNPLANNED rise in containment pressure Exceeding one or more Auxiliary Building Control MAX SAFE area radiation levels (EP-4)

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 85 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument which is re-spanned to indicate water level in the refuel cavity and the Core Plate d/p instrument which is re-spanned and re-scaled to indicate water level. (ref. 1).

In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in Suppression Pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory.

In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise, with corresponding indications on area radiation monitors. 100R/hr is used for this indication on Containment High Range Radiation Monitors (1D21-K648B and C. These detectors are located on the containment wall in a position to monitor the containment radiation environment above the refueling cavity elevation.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 86 of 245 Attachment 2 Emergency Action Level Technical Bases Four conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE is not established.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment.

However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%).

The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 4).

Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.

Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control, (ref. 5).

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 87 of 245 Attachment 2 Emergency Action Level Technical Bases damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Reference(s):

1. 03-1-01-5 Refueling

2. 04-1-02-1H13-P601 Alarm Response Instruction Panel 1H13-P601

3. 04-1-02-1H13-P680 Alarm Response Instruction Panel 1H13-P680

4. BWROG Emergency Procedure and Severe Accident Guidelines, Revision 3, p. B-16-64

5. 05-S-01-EP-4, Auxiliary Building Control

6. 06-IC-1D21-R-1002 Containment/Drywell High Range Area Radiation Monitor Calibration

7. NEI 99-01 CG1

8. Calculation J-D21-1, Set Points Determination for High Range DW & Containment Radiation Monitors (D21 System)

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 88 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of ESF AC Power Initiating Condition: Loss of all but one AC power source to ESF buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-3, to DIV I and DIV II ESF 4.16 KV buses reduced to a single power source for 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table C-3 AC Power Sources Offsite ESF Transformer 11 ESF Transformer 12 ESF Transformer 21 Onsite DIV I DG (DG 11)

DIV II DG (DG 12)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 89 of 245 Attachment 2 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

The HPCS bus (DIV III) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the greater time available to restore another power source to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency ESF power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being back-fed from the unit main generator.

A loss of emergency ESF power sources (e.g., onsite diesel generators) with a single train of emergency ESF buses being fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

This EAL is the cold condition equivalent of the hot condition EAL SA1.1.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 90 of 245 Attachment 2 Emergency Action Level Technical Bases

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. NEI 99-01 CU2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 91 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of ESF AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4.16 KV buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

Mitigative strategies using other power sources (HPCS DIV III diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, available power for suppression pool cooling systems is appropriate for this EAL.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 92 of 245 Attachment 2 Emergency Action Level Technical Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore an ESF bus to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SS1.1.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. NEI 99-01 CU2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 93 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED rise in RCS temperature to > 200°F Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F) (ref. 1, 2). In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification is based on the concurrent loss of reactor vessel level indications per EAL CU3.2.

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 94 of 245 Attachment 2 Emergency Action Level Technical Bases During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. Technical Specifications Table 1.1-1

2. 03-1-01-3 Plant Shutdown

3. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal

4. NEI 99-01 CU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 95 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

4 - Cold Shutdown, 5- Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to EAL CA3.1.

This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 96 of 245 Attachment 2 Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Reference(s):

1. 02-S-01-40 EP Technical Bases

2. Technical Specifications Table 1.1-1

3. 03-1-01-3 Plant Shutdown

4. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal

5. NEI 99-01 CU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 97 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED rise in RCS temperature to > 200°F for > Table C-4 duration (Note 1)

OR UNPLANNED RPV pressure rise > 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table C-4 RCS Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status Intact N/A 60 min.*

established 20 min.*

Not intact not established 0 min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The actions taken to secure containment (Primary or Secondary) and their associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when either Primary or Secondary Containment integrity is established.

UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 98 of 245 Attachment 2 Emergency Action Level Technical Bases Basis:

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4.

This EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.

The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantial degradation in plant safety.

Finally, in the case where there is a rise in RCS temperature, the RCS is not intact and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).

This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The RCS pressure rise threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or AS1.

Reference(s):

1 Technical Specifications Table 1.1-1

2. 03-1-01-3 Plant Shutdown

3. 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat Removal

4. NEI 99-01 CA3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 99 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event Indicated voltage is < 105 VDC on required vital 125 VDC buses 11DA and 11DB for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis Vital DC buses 11DA and 11DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1.72 volts/cell.

These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC)

(ref. 1, 2).

This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions raise the time available to restore a J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 100 of 245 Attachment 2 Emergency Action Level Technical Bases vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, required means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification because the other vital bus remains powered supporting in-service equipment.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category A.

This EAL is the cold condition equivalent of the hot condition EAL SS2.1.

Reference(s):

1. Calculation No: EC-Q1111-14001 Station Division I Battery 1A3 and Division II Battery 1B3 Discharge Capacity during Extended Loss of AC Power

2. UFSAR 8.3.2.1.1 Station DC Power

3. NEI 99-01 CU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 101 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods State/

System Onsite NRC Local Station Radio System X GGNS Plant Phone System X Public Address System X Emergency Notification System (ENS) X Commercial Telephone System X X Satellite Phones X X Operational Hotline X Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, DEF - Defueled Definition(s):

None Basis:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 102 of 245 Attachment 2 Emergency Action Level Technical Bases This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriffs Department, Louisiana Department of Environmental Quality, Tensas Parish Sheriffs Office, and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness.

The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

Reference(s):

1. GGNS Emergency Plan Section 7.5, Communications Systems

2. 04-S-01-R61-1 Plant Communications

3. NEI 99-01 CU5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 103 of 245 Attachment 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table C-6 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 104 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 105 of 245 Attachment 2 Emergency Action Level Technical Bases such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC CS1 or AS1.

This EAL is the cold condition equivalent of the hot condition EAL SA8.1.

Reference(s):

1. EP FAQ 2016-002

2. NEI 99-01 CA6 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 106 of 245 Attachment 2 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

The GGNS ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 107 of 245 Attachment 2 Emergency Action Level Technical Bases Category: E - ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack)

> EITHER of the following:

60 mrem/hr (gamma + neutron) on the top of the overpack 600 mrem/hr (gamma + neutron) on the side of the overpack (excluding inlet and outlet ducts)

Mode Applicability:

All Definition(s):

CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the GGNS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC).

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of damage is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values. The technical specification (licensing bases document) multiple of 2 times, which is also used in Recognition Category A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 108 of 245 Attachment 2 Emergency Action Level Technical Bases IC AU1, is used here to distinguish between non-emergency and emergency conditions (ref. 2).

The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference(s):

1. UFSAR 9.1.4.2.10.4 Storage of Fuel at the Independent Spent Fuel Storage Installation

2. GGNS HI-STORM 100 10 CFR 72.212 Evaluation Report Licensing Basis Document, Revision 10, Section 4.2.4 (Section 5.7) Radiation Protection Program

3. NEI 99-01 E-HU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 109 of 245 Attachment 2 Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves.

C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. Loss means the barrier no longer assures containment of radioactive materials. Potential Loss means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:

Alert:

Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 110 of 245 Attachment 2 Emergency Action Level Technical Bases Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded.

The fission product barrier thresholds specified within a scheme reflect plant-specific GGNS design and operating characteristics.

As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.

The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.

At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 111 of 245 Attachment 2 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 Reference(s):

1. NEI 99-01 FA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 112 of 245 Attachment 2 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

One barrier loss and a second barrier loss (i.e., loss - loss)

One barrier loss and a second barrier potential loss (i.e., loss - potential loss)

One barrier potential loss and a second barrier potential loss (i.e., potential loss -

potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT.

Reference(s):

1. NEI 99-01 FS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 113 of 245 Attachment 2 Emergency Action Level Technical Bases Category: F - Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Containment Barriers Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference(s):

1. NEI 99-01 FG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 114 of 245 Attachment 2 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RPV Water Level B. RCS Leak Rate C. Containment Conditions D. Containment Radiation / RCS Activity E. Containment Integrity or Bypass F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word None is entered in the cell.

Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2FCB6).

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 115 of 245 Attachment 2 Emergency Action Level Technical Bases potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then B,, F.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 116 of 245 Attachment 2 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB)

Category Loss Potential Loss Loss Potential Loss Loss Potential Loss FCB2 RPV water level cannot A be restored and RCB1 RPV water level cannot be restored and maintained FCB1 SAP entry is required maintained None None CNB1 SAP entry is required RPV Water > -167 in. (TAF)

> -167 in. (TAF)

Level or cannot be determined or cannot be determined RCB2 UNISOLABLE break in any RCB4 UNISOLABLE primary system CNB2 UNISOLABLE primary of the following: leakage that results in system leakage that results Main steam line exceeding EITHER: in exceeding EITHER:

B None None RCIC steam Line RWCU One or more EP-4 radiation Operating Limits One or more EP-4 MAX SAFE area radiation levels None RCS Leak Rate Feedwater HPCS One or more EP-4 area One or more EP-4 MAX temperature Operating SAFE area temperature RCB3 Emergency Depressurization Limits levels is required CNB5 Containment pressure > 15 psig CNB3 UNPLANNED rapid drop in containment pressure CNB6 Drywell or containment hydrogen C RCB5 Drywell pressure > 1.23 psig following containment pressure rise concentration > 4%

None None None CTMT due to RCS leakage CNB4 Containment pressure CNB7 Parameters cannot be restored Conditions response not consistent with and maintained within the safe LOCA conditions zone of the HCTL curve (EP Figure 1)

FCB3 Containment radiation D (RITS-K648B or C) >

400 R/hr RCB6 Drywell radiation (RITS- CNB8 Containment radiation (RITS-CTMT Rad / FCB4 Primary coolant None None None K648A or D) > 100 R/hr K648B or C) > 7000 R/hr RCS activity Activity > 300 µCi/gm dose equivalent I-131 CNB9 UNISOLABLE direct E downstream pathway to the environment exists after CTMT None None None None Containment isolation signal None Integrity or Bypass CNB10 Intentional Containment venting per EPs F FCB5 Any condition in the opinion of the FCB6 Any condition in the opinion of the Emergency RCB7 Any condition in the opinion RCB8 Any condition in the opinion of CNB11 Any condition in the opinion CNB12 Any condition in the opinion of the of the Emergency Director the Emergency Director that of the Emergency Director Emergency Director that indicates Emergency Emergency Director Director that indicates that indicates loss of the indicates potential loss of the that indicates loss of the potential loss of the Containment Director that indicates loss of potential loss of the fuel RCS barrier RCS barrier Containment barrier barrier Judgment the fuel clad barrier clad barrier J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 117 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

FCB1 SAP entry is required Definition(s):

None Basis:

Emergency Procedures (EPs) specify entry to the Severe Accident Procedures (SAPs) when core cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2).

The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.

This threshold is also a Potential Loss of the Containment barrier (CNB1). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCB1). SAP entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

The Loss threshold represents the EOP requirement for entry into the SAPs. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.

Reference(s):

1. 05-S-01-EP-2 RPV Control

2. 05-S-01-EP-5 RPV Flooding

3. EP FAQ 2015-004

4. NEI 99-01, RPV Water Level Fuel Clad Loss 2.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 118 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

FCB2 RPV water level cannot be restored and maintained > -167 in. (TAF) or cannot be determined Definition(s):

None Basis:

An RPV water level instrument reading of -167 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.

When RPV water level cannot be determined, EPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EP-5 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 119 of 245 Attachment 2 Emergency Action Level Technical Bases This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.

The RPV water level threshold is the same as RCS barrier Loss threshold RCB1. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term cannot be restored and maintained above means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

Reference(s):

1. 05-S-01-EP-2 RPV Control

2. 05-S-01-EP-5 RPV Flooding

3. 05-S-01-EP-2A ATWS RPV Control 4 NEI 99-01 RPV Water Level Potential Loss 2.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 120 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 122 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Loss Threshold:

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 124 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

FCB3 Containment radiation (RITS-K648B or C) > 400 R/hr Definition(s):

None Basis:

The containment radiation monitor reading (425 R/hr rounded to 400 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 1.6% fuel clad damage (ref. 1). Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold RCB6 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation / RCS Activity.

Reference(s):

1. XC-Q1D21-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values

2. 04-1-01-D21-1 SOI Area Radiation Monitoring System

3. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 125 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

FCB4 Coolant activity > 300 Ci/gm dose equivalent I-131 Definition(s):

None Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation / RCS Activity.

Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 127 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 128 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 129 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

FCB5 Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad Barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 130 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

FCB6 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad Barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 131 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RPV Water Level Degradation Threat: Loss Threshold:

RCB1 RPV water level cannot be restored and maintained > -167 in. (TAF) or cannot be determined Definition(s):

None.

Basis:

An RPV water level instrument reading of -167 in. indicates level is at the top of active fuel (TAF) (ref. 1). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Containment barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the lowering level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.

When RPV water level cannot be determined, EOPs require entry to EP-5, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). The instructions in EP-5 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss RCB3).

Note that EP-2A, ATWS RPV Control, may require intentionally lowering RPV water level to

-167 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - RPS Failure EALs.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification.

This water level corresponds to the top of active fuel and is used in the EOPs to indicate a challenge to core cooling.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 132 of 245 Attachment 2 Emergency Action Level Technical Bases The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold FCB2.

Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, cannot be restored and maintained above, means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

There is no RCS barrier Potential Loss threshold associated with RPV Water Level.

Reference(s):

1. 05-S-01-EP-2 RPV Control

2. 05-S-01-EP-5 RPV Flooding

3. 05-S-01-EP-2A ATWS RPV Control

4. NEI 99-01 RPV Water Level RCS Loss 2.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 133 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold:

None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 134 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

RCB2 UNISOLABLE break in any of the following:

Main steam line RCIC steam line RWCU Feedwater HPCS Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside containment (e.g., EP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see Loss CNB9) barriers and justifies declaration of a Site Area Emergency (i.e.,

Loss or Potential Loss of any two barriers).

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 135 of 245 Attachment 2 Emergency Action Level Technical Bases Even though the High Pressure Core Spray (HPCS) injects into the RCS, it is included in this EAL due to the potential for an inter-system loss of coolant back flowing from the discharge lines (via failed isolation valves and check valves) and out through a break in the piping. A HPCS failure that does not result in back flow of RCS and out through a break should not be considered as meeting the EAL threshold.

Large high-energy lines that rupture outside containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, remotely or locally,, the RCS barrier Loss threshold is met.

Reference(s):

1. NEI 99-01 RCS Leak Rate RCS Loss 3.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 136 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

RCB3 Emergency Depressurization is required Definition(s):

None Basis:

Emergency Depressurization in accordance with the EOPs (ref. 1, 2) is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

EP-2 RPV Control - Emergency Depressurization allows terminating the depressurization if necessary to maintain RCIC as an injection source. This would require closing the SRVs. Even though the SRVs may be reclosed, this threshold is still met due to the requirement for an Emergency Depressurization having been met (ref. 2).

Reference(s):

1. 05-S-01-EP-2 RPV Control - Emergency Depressurization

2. EP FAQ 2015-003

3. NEI 99-01 RCS Leak Rate RCS Loss 3.B J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EP-4 area radiation Operating Limits One or more EP-4 area temperature Operating Limits Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of UNISOLABLE primary system leakage outside the containment. The EP-4 entry condition values define this RCS threshold because they are the Operating Limits (maximum normal operating values) and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or mis-operation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1).

In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Auxiliary Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 138 of 245 Attachment 2 Emergency Action Level Technical Bases feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment.

Potential loss of RCS based on primary system leakage outside the containment is determined from EOP temperature or radiation EP-4 Operating Limits (Max Normal Operating values) in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside containment.

An EP-4 Operating Limit value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a reduction in the steam or water being discharged through an unisolated break in the system.

An UNISOLABLE leak which is indicated by EP-4 Operating Limit values escalates to a Site Area Emergency when combined with Containment Barrier Loss thresholds CNB 2 or CNB9 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Reference(s):

1. 05-S-01-EP-4 Auxiliary Building Control

2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 139 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: C. CTMT Conditions Degradation Threat: Loss Threshold:

RCB5 Drywell pressure > 1.23 psig due to RCS leakage Definition(s):

None Basis:

The drywell high pressure scram setpoint is an entry condition to EP-2, RPV Control, and EP-3, Containment Control (ref. 1, 2). Normal containment pressure control functions (e.g., operation of drywell and containment cooling, vent using containment vessel purge, etc.) are specified in EP-3 in advance of less desirable but more effective functions (e.g., operation of containment sprays, etc.).

In the design basis, containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the rising pressure trend.

Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control containment vent/purge (ref. 3).

The threshold phrase due to RCS leakage focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect containment pressure. Drywell pressure greater than 1.23 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.23 psig should not be considered an RCS barrier Loss.

The 1.23 psig value is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

There is no RCS barrier Potential Loss threshold associated with CTMT Conditions.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 140 of 245 Attachment 2 Emergency Action Level Technical Bases Reference(s):

1. 05-S-01-EP-2 RPV Control

2. 05-S-01-EP-3 Containment Control

3. UFSAR Section 6.2.1, Containment Functional Design

4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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RCB6 Drywell radiation (RITS-K648A or D) > 100 R/hr Definition(s):

None Basis:

The drywell radiation monitor reading (150 R/hr rounded to 100 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits (ref. 1). This value is lower than that specified for Fuel Clad Barrier Loss threshold FCB3 since it indicates a loss of the RCS Barrier only.

There is no RCS barrier Potential Loss threshold associated with CTMT Radiation/ RCS Activity.

Reference(s):

1. XC-Q1D21-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values

2. 04-1-01-D21-1 SOI Area Radiation Monitoring System

3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 146 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

RCB7 Any condition in the opinion of the Emergency Director that indicates loss of the RCS Barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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RCB8 Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS Barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 148 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Loss Threshold:

None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB1 SAP entry is required Definition(s):

None Basis:

EPs specify entry to the SAPs when core cooling is severely challenged. These EPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2).

The EP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.

This threshold is also a Loss of the Fuel Clad barrier (Loss FCB1). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (Loss RCB1). SAP entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold FCB1. The Potential Loss requirement for entry into the SAGs indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs specify the conditions when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and greater potential for containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

There is no Containment barrier Loss threshold associated with RPV Water Level.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 150 of 245 Attachment 2 Emergency Action Level Technical Bases Reference(s):

1. 05-S-01-EP-2 RPV Control

2. 05-S-01-EP-5 RPV Flooding

3. EP FAQ 2015-004

4. NEI 99-01 RPV Water Level PC Potential Loss 2.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 151 of 245 Attachment 2 Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EP-4 MAX SAFE area radiation levels One or more EP-4 MAX SAFE area temperature levels Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the containment. The MAX SAFE values define this Containment barrier threshold because they are indicative of problems in the Secondary Containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EP-4, Auxiliary Building Control (ref. 1).

In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area temperature condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by a fire or loss of area cooling. Conversely, a high area temperature condition in conjunction with other indications (e.g. room FLOODING, high area radiation levels, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 152 of 245 Attachment 2 Emergency Action Level Technical Bases The Max Safe area temperature values and the Max Safe area radiation values are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

There is no Containment barrier Potential Loss threshold associated with RCS Leak Rate.

Reference(s):

1. 05-S-01-EP-4 Auxiliary Building Control

1. NEI 99-01 RCS Leak Rate PC Loss 3.C J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB3 UNPLANNED rapid drop in containment pressure following containment pressure rise Definition(s):

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Rapid UNPLANNED loss of containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity.

This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB4 Containment pressure response not consistent with LOCA conditions Definition(s):

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Containment pressure should rise as a result of mass and energy release into the containment from a LOCA. Thus, containment pressure not rising under these conditions indicates a loss of containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

Reference(s):

1. USAR Table 6.2-5, Summary of Short Term Containment Responses to Recirculation Line and Main Steam Line Breaks

2. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design

3. NEI 99-01 Primary Containment Conditions PC Loss 1.B J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB5 Containment pressure > 15 psig Definition(s):

None Basis:

When the containment pressure exceeds the maximum allowable value (15 psig) (ref. 1),

containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). This pressure is based on the containment design pressure as identified in the accident analysis. If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.

The threshold pressure is the containment internal design pressure. Structural acceptance testing demonstrates the capability of the containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

Reference(s):

1. UFSAR Table 6.2-13, Maximum Calculated Accident for Containment Design

2. 05-S-01-EP-3 Containment Control

3. NEI 99-01, Primary Containment Conditions PC Potential Loss 1.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB6 Drywell or containment hydrogen concentration > 4%

Definition(s):

None Basis:

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition.

Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 1).

If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the containment, loss of the Containment barrier could occur.

Reference(s):

1. 02-S-01-40 EP Technical Bases (EP-3 step H-3)

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.B J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB7 Parameters cannot be restored and maintained within the safe zone of the HCTL curve (EP Figure 1)

Definition(s):

None Basis:

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

The term cannot be restored and maintained within means the parameter value(s) is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to the parameter value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained within a specified limit does not require immediate action simply because the current value is outside the limit, but does not permit extended operation outside the limit; the threshold must be considered reached as soon as it is apparent that operation within the limit cannot be attained.

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1. 05-S-01-EP-3 Containment Control

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB8 Containment radiation (RITS-K648B or C) > 7,000 R/hr Definition(s):

None Basis:

In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (Loss RCB6) and a loss of the Fuel Clad barrier (Loss FCB3) have already occurred. This threshold, therefore, represents a General Emergency classification.

The containment radiation monitor reading (7,350 R/hr rounded to 7,000 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed (ref. 1). This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

There is no Containment barrier Loss threshold associated with CTMT Radiation/RCS Activity.

Reference(s):

1. XC-Q1D21-17001 Grand Gulf Nuclear Station (GGNS) Containment Radiation EAL Threshold Values

2. 04-1-01-D21-1 SOI Area Radiation Monitoring System

3. NEI 99-01 NEI 99-01 Primary Containment Radiation Potential Loss 4.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB9 UNISOLABLE direct downstream pathway to the environment exists after Containment isolation signal Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity.

This threshold also applies to a containment bypass due to a HPCS or LPCS line break outside containment with injection check valve failure allowing an UNISOLABLE direct pathway for RCS release to the environment.

The use of the modifier direct in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE main steam line or RCIC steam line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the main condenser is available with an UNISOLABLE main steam line, there may J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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/ Rad Effluent, EALs.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

EP-3 Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a VALID containment isolation signal, the Containment barrier should be considered lost.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system.

These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.

There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass.

Reference(s):

1. 05-S-01-EP-3 Containment Control

2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB10 Intentional Containment venting per EPs Definition(s):

None Basis:

EP-3, Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded.

The threshold is met when the operator begins venting the containment in accordance with 3, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 1).

Intentional venting of containment for containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.

There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass.

Reference(s):

1. 05-S-01-EP-3 Containment Control

2. NEI 99-01 CTMT Integrity or Bypass Containment Loss 3.B J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB11 Any condition in the opinion of the Emergency Director that indicates loss of the Containment Barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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CNB12 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment Barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 168 of 245 Attachment 2 Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.

1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown

5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 170 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by GGNS Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 171 of 245 Attachment 2 Emergency Action Level Technical Bases SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 172 of 245 Attachment 2 Emergency Action Level Technical Bases The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Security Plan for GGNS.

The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with 11-S-82-1 Security Contingency Events (ref. 2).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1).

Escalation of the emergency classification level would be via IC HA1.

Reference(s):

1. GGNS Security Plan

2. 11-S-82-1 Security Contingency Events

3. NEI 99-01 HU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 173 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the SECURITY OWNER CONTROLLED AREA as reported by GGNS Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 174 of 245 Attachment 2 Emergency Action Level Technical Bases SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the SECURITY OWNER CONTROLLED AREA.

The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with 11-S-82-1 Security Contingency Events (ref. 2).

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 175 of 245 Attachment 2 Emergency Action Level Technical Bases In some cases, it may not be readily apparent if an aircraft impact within the SECURITY OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1).

Escalation of the emergency classification level would be via IC HS1.

Reference(s):

1. GGNS Security Plan

2. 11-S-82-1 Security Contingency Events

3. NEI 99-01 HA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 176 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by GGNS Security Shift Supervision Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 177 of 245 Attachment 2 Emergency Action Level Technical Bases Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2).

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for GGNS (ref. 1).

Reference(s):

1. GGNS Security Plan

2. 11-S-82-1 Security Contingency Events

3. NEI 99-01 HS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 178 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL:

HU2.1 Unusual Event Seismic event > OBE as indicated by annunciation of EITHER of the following on SH13P856:

Containment Operating Basis Earthquake (P856-1A-A3)

Drywell Operating Basis Earthquake (P856-1A-A5)

Mode Applicability:

All Definition(s):

None Basis:

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., perform walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the U.S. Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 179 of 245 Attachment 2 Emergency Action Level Technical Bases To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst with the following GGNS coordinates: 32º 0' 27" north latitude, 91º 2' 53" west longitude (ref. 2).

Alternatively, near real-time seismic activity can be accessed via the NEIC website.

Reference(s):

1. 05-S-02-Vl-3 Earthquake

2. UFSAR 2.1.1 Site Location and Description

3. NEI 99-01 HU2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 180 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a tornado striking (touching down) within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1.

A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Reference(s):

1. 05-1-02-VI-2 Hurricanes, Tornados and Severe Weather

2. NEI 99-01 HU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 181 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability:

All Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 182 of 245 Attachment 2 Emergency Action Level Technical Bases warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Refer to EAL CA6.1 or SA8.1 for internal FLOODING affecting more than one SAFETY SYSTEM train.

Reference(s):

1. 05-1-02-VI-1 Flooding

2. NEI 99-01 HU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 183 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s):

1. NEI 99-01 HU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 184 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, S or C.

Reference(s):

1. NEI 99-01 HU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 185 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table H-1 Fire Areas Unit 1 Containment Unit 1 Auxiliary Building Unit 1 Turbine Building Control Building Diesel Generator Rooms SSW Pump & Valve Rooms Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 186 of 245 Attachment 2 Emergency Action Level Technical Bases VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.

Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

Reference(s):

1. 05-S-02-V-1 Response to Fires

2. 10-S-03-2 Response to Fires

3. NEI 99-01 HU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 187 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) (Note 11)

AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 11: During Modes 1 and 2, HU4.2 is not applicable to a single fire alarm in the containment or drywell.

Table H-1 Fire Areas Unit 1 Containment Unit 1 Auxiliary Building Unit 1 Turbine Building Control Building Diesel Generator Rooms SSW Pump & Valve Rooms Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 188 of 245 Attachment 2 Emergency Action Level Technical Bases VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

This EAL is not applicable for the containment or drywell in Modes 1 and 2. The air flow design and TS requirements for operation of Containment Fan Coolers and the drywell cooling system are such that multiple detectors would be expected to alarm for a fire in the containment or drywell. A fire in the containment or drywell in these modes would therefore be classified under EAL HU4.1 If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report.

If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 189 of 245 Attachment 2 Emergency Action Level Technical Bases Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

The 30 minute requirement begins upon receipt of a single VALID fire detection system alarm.

The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15 minute requirement beginning with the verification of the fire by field report.

Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).

Reference(s):

1. 05-S-02-V-1 Response to Fires

2. 10-S-03-2 Response to Fires

3. UFSAR Appendix 9A Fire Hazard Analysis Report

4. NEI 99-01 HU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 190 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

Reference(s):

1. NEI 99-01 HU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 191 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1.

Reference(s):

1. NEI 99-01 HU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 192 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Control Building 111 SWGR Rms (0C202, 0C215) 3 Auxiliary Building 93 RHR A Pump Room (1A103) 3 Auxiliary Building 93 RHR B Pump Room (1A105) 3 Auxiliary Building 93 Corridor (1A101) 3 Auxiliary Building 119 Corridor (1A201) 3 Auxiliary Building 119 RHR A Pump Room (1A203) 3 Auxiliary Building 119 RHR B Pump Room (1A205) 3 Auxiliary Building 119 RCIC Room (1A204) 3 Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 Radwaste Building 118 Radwaste Control Room (0R241) 3 Mode Applicability:

3 - Hot Shutdown J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 193 of 245 Attachment 2 Emergency Action Level Technical Bases Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Directors judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if any of the following conditions apply:

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.

The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 194 of 245 Attachment 2 Emergency Action Level Technical Bases An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.

Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area.

Escalation of the emergency classification level would be via Recognition Category A, C or F ICs.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

EAL HA5.1 mode applicability has been limited to the mode limitations of Table H-2 (Mode 3 only).

Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases

2. NEI 99-01 HA5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 195 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel Mode Applicability:

All Definition(s):

None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Transfer of plant control begins when the last licensed operator leaves the Control Room.

Escalation of the emergency classification level would be via IC HS6.

Reference(s):

1. 05-1-02-II-1 Shutdown from the Remote Shutdown Panel

2. NEI 99-01 HA6 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 196 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel AND Control of any of the following key safety functions is not re-established within 15 min.

(Note 1):

Reactivity (Modes 1 and 2 only)

RPV water level RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling Definition(s):

None Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 197 of 245 Attachment 2 Emergency Action Level Technical Bases Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room.

Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):

1. 05-1-02-II-1 Shutdown from the Remote Shutdown Panel

2. EP FAQ 2015-014

3. NEI 99-01 HS6 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 198 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Mode Applicability:

All Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT.

Reference(s):

1. NEI 99-01 HU7 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 199 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 200 of 245 Attachment 2 Emergency Action Level Technical Bases Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT.

Reference(s):

1. NEI 99-01 HA7 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 201 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL:

HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

SITE BOUNDARY - That boundary defined by a 696 meter (.43 miles) radius from the center of the reactor.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 202 of 245 Attachment 2 Emergency Action Level Technical Bases SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY.

Reference(s):

1. NEI 99-01 HS7 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 203 of 245 Attachment 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL:

HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Mode Applicability:

All Definition(s):

HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward GGNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on GGNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SECURITY OWNER CONTROLLED AREA).

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 204 of 245 Attachment 2 Emergency Action Level Technical Bases PROTECTED AREA - An area encompassed by physical barriers (i.e., the security fence) and to which access is controlled.

SECURITY OWNER CONTROLLED AREA (SOCA) - The SOCA is the area demarcated as a Vehicle Barrier System (VBS) consisting of passive elements including a series of large concrete blocks on the inside of a delay fence with early warning capabilities. The SOCA is the area between the SOCA Fence and the PROTECTED AREA boundary Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY.

Reference(s):

1. NEI 99-01 HG7 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 205 of 245 Attachment 2 Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of ESF AC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV ESF buses.

2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.

3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core.

The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 206 of 245 Attachment 2 Emergency Action Level Technical Bases indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 207 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite AC power capability to ESF buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 KV buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table S-1 AC Power Sources Offsite ESF Transformer 11 ESF Transformer 12 ESF Transformer 21 Onsite DIV I DG (DG 11)

DIV II DG (DG 12)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The HPCS bus (DIV III) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout.

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC ESF buses. This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, capability means that an offsite AC power source(s) is available to the ESF buses, whether or not the buses are powered from it.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 208 of 245 Attachment 2 Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC SA1.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. NEI 99-01 SU1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 209 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all but one AC power source to ESF buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to DIV I and DIV II ESF 4.16 KV buses reduced to a single power source for 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table S-1 AC Power Sources Offsite ESF Transformer 11 ESF Transformer 12 ESF Transformer 21 Onsite DIV I DG (DG 11)

DIV II DG (DG 12)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 210 of 245 Attachment 2 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

The HPCS bus (DIV III) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1.

An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an ESF bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all ESF emergency power sources (e.g., onsite diesel generators) with a single train of ESF buses being back-fed from the unit main generator.

A loss of ESF emergency power sources (e.g., onsite diesel generators) with a single train of ESF emergency buses being fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC SS1.

This EAL is the hot condition equivalent of the cold condition EAL CU2.1.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. NEI 99-01 SA1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 211 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to ESF buses for 15 minutes or longer EAL:

SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4.16 KV buses for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

Mitigative strategies using other power sources (HPCS DIV III diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 212 of 245 Attachment 2 Emergency Action Level Technical Bases particular, available power for suppression pool cooling systems is appropriate for this EAL. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC AG1, FG1 or SG1.

This EAL is the hot condition equivalent of the cold condition EAL CA2.1.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. NEI 99-01 SS1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 213 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to ESF buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4.16 KV buses AND EITHER:

Restoration of at least one ESF 4.16 KV bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

RPV water level cannot be restored and maintained > -191 in.

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-191 in.)

(ref. 6). Core submergence is the most desirable means of core cooling, however when RPV J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 214 of 245 Attachment 2 Emergency Action Level Technical Bases level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e.,

steam cooling or spray cooling).

This IC addresses a prolonged loss of all power sources to AC ESF emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (HPCS DIV III diesel generator, FLEX generators, etc.)

may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, available power for suppression pool cooling systems is appropriate for this EAL. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC ESF emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is a greater likelihood of challenges to multiple fission product barriers. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the site-specific SBO coping analysis time (ref. 4).

The estimate for restoring at least one ESF emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. 02-S-01-40 EP Technical Bases

7. NEI 99-01 SG1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 215 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of ESF AC Power Initiating Condition: Loss of all ESF AC and vital DC power sources for 15 minutes or longer EAL:

SG1.2 General Emergency Loss of all offsite and all onsite AC power to DIV I and DIV II ESF 4.16 KV buses for 15 min. (Note 1)

AND Indicated voltage is < 105 VDC on vital 125 VDC buses 11DA and 11DB for 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Vital DC buses 11DA and 111DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1.72 volts/cell.

These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC)

(ref. 6, 7).

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 216 of 245 Attachment 2 Emergency Action Level Technical Bases This IC addresses a concurrent and prolonged loss of both emergency ESF AC and Vital DC power. A loss of all emergency ESF AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

Mitigative strategies using other power sources (HPCS DIV III diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, available power for suppression pool cooling systems is appropriate for this EAL. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency ESF AC and Vital DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

Reference(s):

1. UFSAR Figure 8.1-001 Main One Line Diagram

2. UFSAR section 8.1 Electric Power Introduction

3. UFSAR section 8.3 Onsite Power

4. UFSAR section 8A Loss of all AC Power

5. 05-1-02-I-4 Loss of AC Power

6. Calculation No: EC-Q1111-14001 Station Division I Battery 1A3 and Division II Battery 1B3 Discharge Capacity during Extended Loss of AC Power

7. UFSAR 8.3.2.1.1 Station DC Power

8. NEI 99-01 SG8 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 217 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:

SS2.1 Site Area Emergency Indicated voltage is < 105 VDC on vital 125 VDC buses 11DA and 11DB for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Vital DC buses 11DA and 11DB feed the Division 1 and Division 2 loads respectively. The Division 1 and Division 2 batteries each have 61 cells with a design minimum of 1.72 volts/cell.

These cell voltages yield minimum design bus voltages of 104.92 VDC (rounded to 105 VDC)

(ref. 1, 2).

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 218 of 245 Attachment 2 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AG1, FG1 or SG1.

This EAL is the hot condition equivalent of the cold condition EAL CU4.1.

Reference(s):

1. Calculation No: EC-Q1111-14001 Station Division I Battery 1A3 and Division II Battery 1B3 Discharge Capacity during Extended Loss of AC Power

2. UFSAR 8.3.2.1.1 Station DC Power

3. NEI 99-01 SS8 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 219 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table S-2 Safety System Parameters Reactor power RPV water level RPV pressure Containment pressure Suppression Pool water level Suppression Pool temperature Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 220 of 245 Attachment 2 Emergency Action Level Technical Bases UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via EAL SA3.1.

Reference(s):

1. UFSAR 7.5 Safety-Related Display Instrumentation

2. NEI 99-01 SU2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 221 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Table S-2 Safety System Parameters Reactor power RPV water level RPV pressure Containment pressure Suppression Pool water level Suppression Pool temperature Table S-3 Significant Transients Reactor scram UNPLANNED drop in reactor thermal power > 25%

Electrical load rejection > 25%

electrical load ECCS injection Thermal power oscillations > 10%

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 222 of 245 Attachment 2 Emergency Action Level Technical Bases Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 223 of 245 Attachment 2 Emergency Action Level Technical Bases or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC FS1 or AS1.

Reference(s):

1. UFSAR 7.3 Engineered Safety Features Systems

2. NEI 99-01 SA2 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 224 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Offgas Pretreatment radiation monitor high-high alarm Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The Offgas Pretreatment monitors radioactivity in the Offgas system downstream of the Offgas condenser. The monitor detects the radiation level that is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser. The Hi-Hi alarm, if alarming, indicates that the radioactivity present at the recombiner effluent discharge is at or above the Technical Specification 3.7.5 limit of 380 millicuries per second of Noble Gases. (ref. 1)

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs.

In the event that the Offgas Pretreatment Radiation Monitor High-High alarm is out of service, the use of Offgas flowrates and Offgas Pretreatment Radiation monitor readings is a viable contingency action to classify the EAL. See chart in 04-1-02-1H13-P601-19A-D7 Alarm Response Instruction for OG PRE-TREAT RAD HI_HI alarm.

Reference(s):

1. Alarm Response Instruction 04-1-02-1H13-P601-19A-D7

2. UFSAR 11.5 Process and Effluent Radiological Monitoring and Sampling Systems J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 225 of 245 Attachment 2 Emergency Action Level Technical Bases

3. Technical Specification 3.7.5 Main Condenser Offgas

4. 05-1-02-II-2 Offgas Activity High

5. NEI 99-01 SU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 226 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.2 Unusual Event Coolant activity > 0.2 Ci/gm dose equivalent I-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR Coolant activity > 4.0 Ci/gm dose equivalent I-131 instantaneous Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs.

Reference(s):

1. Technical Specification B3.4.8, RCS Specific Activity bases

2. UFSAR Section 15.6.4 Steam System Piping Break Outside Containment

3. NEI 99-01 SU3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 227 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1)

OR RCS identified leakage > 25 gpm for 15 min. (Note 1)

OR Leakage from the RCS to a location outside containment > 25 gpm for 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a collecting sump; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage.

Unidentified leakage is all leakage into the drywell that is not identified leakage (ref. 2, 3).

Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2, 3).

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 228 of 245 Attachment 2 Emergency Action Level Technical Bases This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage", "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, or a location outside of containment.

The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category A or F.

Reference(s):

1. UFSAR Section 5.2.5, Detection of Leakage Through Reactor Coolant Pressure Boundary

2. Technical Specification Definitions Section 1.1

3. Technical Specification 3.4.5

2. NEI 99-01 SU4 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 229 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic scram did not shut down the reactor as indicated by reactor power > 5%

after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in shutting down the reactor as indicated by reactor power 5% (APRM downscale) (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation, 2 - Startup Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).

A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 230 of 245 Attachment 2 Emergency Action Level Technical Bases For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2).

Following any automatic RPS scram signal, operating procedures (e.g., EP-2) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event (ref. 3).

Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated.

For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to 5% is not considered a successful automatic scram. If automatic initiation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power.

However, a successful automatic or manual initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.

In the event that the operator identifies a reactor scram is IMMINENT and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss.

If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.

Following the failure of an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 231 of 245 Attachment 2 Emergency Action Level Technical Bases subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles. Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via EAL SA6.1. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.

If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. 05-S-01-EP-2A ATWS RPV Control

3. 05-S-01-EP-2 RPV Control

4. NEI 99-01 SU5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 232 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:

SU6.2 Unusual Event A manual scram did not shut down the reactor as indicated by reactor power > 5% after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) is successful in shutting down the reactor as indicated by reactor power 5% (APRM downscale) (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation, 2 - Startup Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 233 of 245 Attachment 2 Emergency Action Level Technical Bases This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power 5%) (ref. 1).

A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%.

For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2).

Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated.

Successful automatic or manual initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.

If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design ( 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1.

Following the failure of an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch. Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 234 of 245 Attachment 2 Emergency Action Level Technical Bases Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.

If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. 05-S-01-EP-SA ATWS RPV Control

3. 05-S-01-EP-2 RPV Control

4. NEI 99-01 SU5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 235 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:

SA6.1 Alert An automatic or manual scram fails to shut down the reactor as indicated by reactor power

> 5%

AND Manual scram actions taken at the reactor control console (Mode Switch, Manual PBs, ARI/RPT) are not successful in shutting down the reactor as indicated by reactor power >

5% (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation, 2 - Startup Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 236 of 245 Attachment 2 Emergency Action Level Technical Bases This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by subsequent manual scram actions that fail to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (> 5%).

For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual scram pushbuttons, or ARI/RPT initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EP-2A step Q-1) does not constitute a successful manual scram (ref. 2).

For the purposes of this EAL, a successful automatic initiation of ARI/RPT that reduces reactor power to or below 5% is not considered a successful automatic scram. If automatic actuation of ARI/RPT has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI/RPT is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power.

However, a successful automatic initiation of ARI/RPT is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.

A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 237 of 245 Attachment 2 Emergency Action Level Technical Bases It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. 05-S-01-EP-2A ATWS RPV Control

3. 05-S-01-EP-2 RPV Control

4. NEI 99-01 SA5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 238 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL:

SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor as indicated by reactor power

> 5%

AND All actions to shut down the reactor are not successful as indicated by reactor power > 5%

AND EITHER:

RPV water level cannot be restored and maintained > -191 in.

OR Heat Capacity Temperature Limit (HCTL) exceeded (EP Figure 1)

Mode Applicability:

1 - Power Operation, 2 - Startup Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 239 of 245 Attachment 2 Emergency Action Level Technical Bases Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

This EAL addresses the following:

Any automatic reactor scram signal followed by subsequent manual scram actions that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1), and Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

Reactor shutdown achieved by use of control rod insertion methods in EP-2A step Q-1 are also credited as a successful shutdown provided reactor power can be reduced to or below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1)

The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.

Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref.

1). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence (ref 2).

The Heat Capacity Temperature Limit (HCTL, EP Figure 1) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppression pool temperature.

The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step of section SPT in EP-3, Containment Control, is reached (ref. 3). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature.

In some instances, the emergency classification resulting from this EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 240 of 245 Attachment 2 Emergency Action Level Technical Bases Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

Escalation of the emergency classification level would be via IC AG1 or FG1.

Reference(s):

1. 05-S-01-EP-2A, ATWS RPV Control

2. 05-S-01-EP-5, RPV Flooding

3. 05-S-01-EP-3, Containment Control

4. NEI 99-01 SS5 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 241 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods State/

System Onsite NRC Local Station Radio System X GGNS Plant Phone System X Public Address System X Emergency Notification System (ENS) X Commercial Telephone System X X Satellite Phones X X Operational Hotline X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

None J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 242 of 245 Attachment 2 Emergency Action Level Technical Bases Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Mississippi Emergency Management Agency, Claiborne County Civil Defense, Mississippi Highway Safety Patrol, Claiborne County Sheriffs Department, Louisiana Department of Environmental Quality, Tensas Parish Sheriffs Office, and the Louisiana Governor's Office of Homeland Security and Emergency Preparedness.

The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

Reference(s):

1. GGNS Emergency Plan Section 7.5, Communications Systems

2. 04-S-01-R61-1 Plant Communications

3. NEI 99-01 SU6 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 243 of 245 Attachment 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

SA8.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table S-5 Hazardous Events Seismic event (earthquake)

Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 244 of 245 Attachment 2 Emergency Action Level Technical Bases Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s):

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 2 Page 245 of 245 Attachment 2 Emergency Action Level Technical Bases words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues.

Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or AS1.

This EAL is the hot condition equivalent of the cold condition EAL CA6.1.

Reference(s):

1. EP FAQ 2016-002

2. NEI 99-01 SA9 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 1 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases

Background

NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.

These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:

The site-specific list of plant rooms or areas with entry-related mode applicability identified should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.

Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 2 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GGNS Table A-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:

IOI / SOI ACTIONS LOCATION MODE NOTES IOI 03-1-01-2 Power Operations LOWER Power by reducing Recirculation flow until 62.2% core MCR 1 flow (70 mlbm/hr) is reached.

INSERT Control Rods per Control Rod Movement Sequence. MCR 1 TECH SPEC TRIGGER (SR 3.3.2.1.2, SR 3.3.2.1.4) MCR 1 IF Reactor power has been reduced below the HPSP OR the LPSP, THEN PERFORM one of the following: Required Surveillances or enter LCO for TS 3.3.2.1 CHECK OPEN the following valves on 1H13-P870-6C: MCR 1

a. N11-F029A, HP TURB EXTR To MSR A 1ST STG RHT

b. N11-F029B, HP TURB EXTR To MSR B 1ST STG RHTIF N11-F029A OR N11-F029B are NOT open, THEN RETURN MSR 1ST Stage Reheaters to service per SOI 04-1-01-N11-1.

CHECK OPEN the following valves on panel 1H13-P870-6C: MCR 1

a. N36-F010A, EXTR STM SPLY TO FW HTR 5A

b. N36-F010B, EXTR STM SPLY TO FW HTR 5B

c. N36-F011A, EXTR STM SPLY TO FW HTR 6A

d. N36-F011B, EXTR STM SPLY TO FW HTR 6B TAKE handswitches for the following valves to OPEN position on panel 1H13-P870-6C:

a. N36-F013A, FW HTR 5A EXTR STM BTV

b. N36-F013B, FW HTR 5B EXTR STM BTV

c. N36-F012A, FW HTR 6A EXTR STM BTV

d. N36-F012B, FW HTR 6B EXTR STM BTV NOTIFY the following of the power reduction: MCR 1

• Load Dispatcher (Woodlands)

• *Duty Manager (IF unexpected power reduction)

• (SMEPA)(1-601-261-2318 OR 1-601-261-2313)

• * (SMEPA) Site Representative

• Radwaste

• Radiation Protection

• Chemistry

• *NRC Resident Inspector

• These notifications Must be made by Shift Manager J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 3 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES IF LP Turbine inlet temperature is >491°F, and N11-F028A and TURB 1 Not N11-F028B are open, THEN SIMULTANEOUSLY THROTTLE BLDG ELEV Required the following valves on 1H22-P177 to CONTROL LP Inlet 133 AREA 4 Temperatures within a band of 470° F to 490° F while monitoring ROOM LP Turbine Inlet differential temperatures within 30° F 1T325 (comparing A side to B side).

• N11-F028A

• N11-F028B

• IF LP Turbine inlet temperature is >491°F, and N11-F028A and N11-F028B are closed, THEN SLOWLY, SIMULTANEOUSLY LOWER MSR-A/B HTG STM FEED CONT manual setpoint to CONTROL LP Inlet Temperatures within a band of 470° F to 490° F while monitoring LP Turbine Inlet differential temperatures within 30° F (comparing A side to B side).

LOWER Reactor power by INSERTING control rods to specified MCR 1 Control Rod in-sequence position per 17-S-02-400.

At approximately 48% Reactor power, PERFORM the following MCR 1 on panel 1H13-P601.

VERIFY the following valves Open:

B21-F033 INBD MSL DR SOL TO MN CNDSR B21-F069 OTBD MSL DR SOL TO MN CNDSR OPEN B21-F016 At approximately 50% Reactor Power, PERFORM the following: SHUTDOWN 1 Reactor Feed Pump per SOI 04 01-N21-1.

VERIFY RFPT B is operating normally on master controller. MCR 1 RAISE FW MASTER LVL CONT setpoint to approximately 39 MCR 1 TRANSFER the RFPT A SP CONT to MAN. MCR 1 SLOWLY LOWER speed of RFPT A USING RFPT A SP CONT MCR 1 by DEPRESSING the OUT pushbutton. OBSERVE speed of RFPT B raises to maintain RPV water level, OR control it manually FURTHER REDUCE speed of RFPT A using RFPT A SP CONT MCR 1 in MAN until it reaches low speed stop.

TRANSFER speed control of RFPT A to SPEED AUTO by MCR 1 DEPRESSING the OBSERVE the FW AUTO pushbutton extinguishes AND the SPEED AUTO, RAISE, AND LOWER pushbuttons backlight.

FURTHER REDUCE RFPT A speed using RFPT A LOWER MCR 1 pushbutton.

WHEN RFPT A speed reaches 1100 rpm, THEN TRIP RFPT A MCR 1 by DEPRESSING the RFPT A MAN TRIP pushbutton J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 4 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES CHECK F014A, RFP A DISCH VLV starts to close. MCR 1 REOPEN F014A, RFP A DISCH VLV WHEN RFPT A coasts down to zero speed, THEN RESET MCR 1 turning gear by pressing A TURN GEAR OPER RESET pushbutton.

OBSERVE turning gear engages automatically, unless RFPT A is rolling on min flow.

IF turning gear fails to engage, THEN MANUALLY ENGAGE the TURB 1 Not turning gear locally by PRESSING DOWN the manual engaging BLDG ELEV Required lever. 133 AREA 3 ROOM 1T307, 1T309 CHECK OPEN/OPEN the following Drain valves on 1H22-P175: N/A N/A These 1N11-F019A, RFPT A HP IN DR VLV steps are not 1N11-F023A, RFPT A HP IN DR VLV required to 1N11-F018A, RFPT A IP IN DR VLV be 1N11-F021A, RFPT A IP IN DR VLV performed 1N11-F042A, RFPT A IP IN DR VLV to Shut down and 1N33-F021A, RFPT A ABOVE SEAT DR Cooldown 1N33-F022A, RFPT A ABOVE SEAT DR the plant.

1N33-F023A, RFPT A BELOW SEAT DR 1N33-F024A, RFPT A BELOW SEAT DR RETURN FW MASTER LVL CONT setpoint to approximately 36" MCR 1 IF desired, RESET RFPT A trip using the RFPT A TRIP RESET MCR 1 pushbutton SHUTDOWN 1 Circulating Wtr Pump per SOI 04-1-01-N71-1 CHECK that CTCS balls are collected AND system shut down.

DEPRESS the BALL CATCH FLAP CATCH pushbutton on Turb Bldg 1 Not P001A (B) MIMIC AND OBSERVE the flap rotates to the CATCH 113 Area 4 Required position. (1T203)

OBSERVE ball collection starts by a rising number of balls in ball Turb Bldg 1 Not collector tank. 113 Area 4 Required (1T203)

After 10 minutes STOP Ball Recirculation pump by Turb Bldg 1 Not DEPRESSING RECIRC PUMP OFF pushbutton on P001A(B) 113 Area 4 Required MIMIC (1T203)

CLOSE Pump Discharge Valve F323A(B). Turb Bldg 1 Not 113 Area 4 Required J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 5 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES (1T203)

PLACE Screens #1 AND #2 in BACKWASH position by Turb Bldg 1 Not DEPRESSING SCREEN BACKWASH pushbutton on P001A (B) 113 Area 4 Required MIMIC AND OBSERVE screens rotate to BACKWASH position. (1T203)

PRESS the CIRC WTR PMP A(B) STOP pushbutton on 1H13- MCR 1 P680.

CHECK that F002A(B) Circulating Water Pump Discharge valve MCR 1 closes on 1H13-P680 ENSURE that A(B) Circulating Water pump has shut down MCR 1 USING pump indication light on 1H13-P680 WHEN its discharge valve is CLOSED.

OPEN OR CHECK OPEN F001 USING CIRC WTR LOOP A/B MCR 1 XTIE handswitch on 1H13-P870.

CLOSE OR CHECK CLOSED F040A (B) Acid Feed Valve. N/A N/A Not required to be performed to Shut down and Cooldown the plant.

CLOSE OR CHECK CLOSED LV-F513 A(B), Blowdown valve MCR 1 OPEN F039A(B), CIRC WTR PUMP A(B) COLUMN VENT N/A N/A Not required to be performed to Shut down and Cooldown the plant.

ENSURE Condenser vacuum is maintained > 23.8" Hg MCR 1 SHUTDOWN one Heater Drain Pump per SOI 04-1-01-N23-1 JOG CLOSED N23-F051A(B), HTR DR PMP A(B) DISCH VLV MCR 1 on 1H13-P680 for desired pump.

STOP HTR DR PMP A(B) on 1H13-P680. MCR 1 WHEN Reactor power has been reduced < 40%,

SHUTDOWN 2nd Heater Drain Pmp per SOI 04-1-01-N23-1 Before securing second Heater Drain Pump, PLACE N23-LK- MCR 1 R053, HTR DR TK DR, in Manual AND Slowly REDUCE output to 0%.

ENSURE Heater Drain Tank level is maintained by Dump Valves MCR 1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 6 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES N23-LV-F518A-E JOG CLOSED N23-F051B(A) HTR DR PMP B(A) DISCH VLV MCR 1 on 1H13-P680 for second pump.

STOP Heater Drain Pump HTR DR PMP B(A) on 1H13-P680. MCR 1 WHEN BOTH Heater Drain Pumps are shutdown, TURB 1 Not THEN CLOSE N23-F054, HTR DR PMP COMMON DISCH VLV BLDG ELEV Required on 1H22 P175 133 AREA 6 ROOM 1T327 SHIFT the Reactor Recirculation Pump(s) to slow speed as MCR 1 follows:

INSERT Control Rods until Load Line is between 50 AND 65%

VERIFY Control Rods are in sequence of the Control Rod Pattern Controller.

BEFORE entering Controlled Entry Region of Figure 3, MCR 1 PERFORM the following WHEN TS 3.3.1.1, Action J.1 is in effect:

VERIFY Fraction of Core Boiling Boundary (FCBB) is 1.0 per 06 RE-1J11-V-0002.

IMPLEMENT TS 3.3.1.1, Action J.2, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entry AND J3 within 90 days.

IF any APRM gain is out of tolerance, THEN ADJUST gain per MCR 1 06-RE-1C51-W-0001 prior to downshift of Recirculation Pumps.

CLOSE Both Recirculation A AND B Flow Control Valves MCR 1 (FCVs) to Min Ed position using RECIRC A(B) FLO CONT on 1H13-P680 TRANSFER Both Reactor Recirculation Pumps to slow speed MCR 1 per SOI 04-1-01-B33-1 CONTINUE Reactor Power reduction to 25 - 30% by insertion of MCR 1 Control Rods SHUTDOWN Hydrogen Water Chemistry Injection per SOI 04-1-01-P73-1.

At H13-P845, momentarily DEPRESS HWC SHUTDOWN MCR 1 pushbutton AND OBSERVE the following:

HWC SHUTDOWN pushbutton 1P73-M602 Will be flashing as H2 AND O2 flows ramp down to 0. O2 isolation valves Will Close WHEN O2 levels remain at normal levels with no O2 injection for at least 5 minutes.

HWC SHUTDOWN pushbutton Will be in solid WHEN all control valves AND isolation valves are fully Closed.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 7 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES HWC RUNNING pushbutton extinguishes.

CLOSE P73-F107, H2 Inj Sply Line Man Line Shutoff valve. N/A N/A Not required to After O2 valves F515 AND F512 (as indicated by white dots on be red cap being perpendicular to pipe) have Closed, CLOSE OR performed CHECK CLOSED Both F207 AND F208, O2 Rack Sply Isol to to Shut OG Preheater A(B).

down and CLOSE 1P73-F209, O2 injection to Condensate pumps Cooldown the plant.

IF Drywell entry is scheduled, WHEN Reactor Power has been N/A N/A Not reduced to less than 30%, THEN PERFORM the following: required to PERFORM the following for 1D21-K607, DRWL PERS HATCH be ARM: performed DIRECT I&C to CONNECT Canon plug to the plug labeled to Shut ALARM AND J3 at the back of 1D21K607. down and Cooldown PLACE Function Selector switch on front of 1D21K607 (DRWL the plant.

PERS HATCH ARM) to OPERATE position.

PERFORM EPI 04-1-03-D21-1 for 1D21K607.

REMOVE Both Second Stage MSR Reheaters from service per SOI 04-1-01-N11-1.

OBSERVE PDS Computer Points N11N044A,B,C AND MCR 1 N11N045A,B, C to monitor LP Turbine Inlet Temperature T during removal of Second Stage Reheaters from service.

ENSURE Both MSR HTG STM FEED CONT are in MANUAL on MCR 1 1H13-P680.

CLOSE the following MSR 2ND STG HTG STM valves on 1H13- MCR 1 P680:

• N11-F304C

• N11-F304D SIMULTANEOUSLY CLOSE the following MSR 2ND STG HTG MCR 1 STM valves on 1H13-P680:

• N11-F304A

• N11-F304B LOWER the manual outputs on Both MSR HTG STM FEED MCR 1 CONT to minimum on 1H13-P680 to close the temperature control valves.

CLOSE the following MSR SUPPLY VLVS valves on 1H22- TURB 1 Not P177. BLDG ELEV Required

• N11- F028A 133 AREA 4

• N11- F028B ROOM 1T325 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 8 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES VERIFY the following valve lineup on local panels: N/A N/A Not

• N35-F015A Closed, HS-M003A required to

• N35-F015B Closed, HS-M003B be performed

• N35-F018A Closed, HS-M007A to Shut

• N35-F018B Closed, HS-M007B down and Cooldown the plant.

IF Feedwater Heater 6A/B are being supplied from extraction TURB 1 Not steam (i.e., IF 1N36-F010A/B AND 1N36-F011A/B on 1H13- BLDG ELEV Required P870 are open), THEN CLOSE the following valves on 1H22- 133 AREA 4 P177: ROOM

• N35-F008A, 2ND STG RHTR DR TK A TO HTR 6A 1T325

• N35-F008B, 2ND STG RHTR DR TK B TO HTR 6B REMOVE Both First Stage MSR Reheaters from service per SOI 04-1-01-N11-1.

OPEN the following valves on 1H13-P870: MCR 1

• N11-F005A, MSR 1ST STG RHT RO BYP DR VLVS

• N11-F005B, MSR 1ST STG RHT RO BYP DR VLVS SIMULTANEOUSLY CLOSE the following valves on 1H13-P870: MCR 1

• N11-F029A, HP TURB EXTR TO MSR A

• N11-F029B, HP TURB EXTR TO MSR B CLOSE the following valves by taking its respective handswitch MCR 1 to TEST:

• N11-F003A, MSR A 1ST STG RHT EXTR STM BTV (1H13-P870)

• N11-F003B, MSR B 1ST STG RHT EXTR STM BTV (1H13-P870)

REMOVE Condensate Precoat filters from service per Not SOI 04-1-01-N22-1, IF in service. required to be performed to Shut down and Cooldown the plant.

OPEN the following BSCV UPSTRM DR VLVs: MCR 1

a. N33-F300A

b. N33-F300B

c. N33-F300C At approximately 23 - 26 % Reactor Power, RAISE the SPEED MCR 1 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 9 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES DEMAND setpoint to approximately 35%, as monitored on PDS Computer point N32K246, by DEPRESSING the SP DEMAND RAISE AND REL pushbuttons.

SIMULTANEOUSLY DEPRESS LOAD REF OFF AND REL MCR 1 pushbuttons on 1H13-P680-9C to turn off load demand Control AND VERIFY OFF light is illuminated.

LOWER load by DEPRESSING SPEED DEMAND LOWER AND MCR 1 REL pushbutton. (Expected value 150 - 175 MWe)

OBTAIN Shift Manager permission for Manual Scram MCR 1/2 NOTIFY the following that Main Generator is being disconnected MCR 1/2 from the grid:

• Entergy Load Dispatcher (Woodlands)

• (SMEPA) 1-601-261-2318 OR 1-601-261-2313)

• Entergy Mississippi Dispatcher

• Duty Manager VERIFY Switchyard lineup is acceptable for trip of J5228 AND MCR 1/2 J5232 INSERT IRMs MCR 1/2 NOTIFY the following personnel/departments that a manual MCR 1/2 scram is being initiated:

• Radwaste

• Chemistry

• Radiation Protection ANNOUNCE over plant pager that manual Scram is being initiated.

TAKE initial temperature data per Attachment III, Data Sheet I of MCR 1/2 IOI 03-1-01-3 prior to scram Manually SCRAM the Reactor using the MANUAL SCRAM MCR 1/2/3 pushbuttons.

a. VERIFY all Control Rods are fully inserted.

b. VERIFY Reactor Power is decreasing.

c. IF Pressure Control System is maintaining reactor pressure greater than 850 psig, THEN PLACE Reactor Mode switch to SHUTDOWN.

d. VERIFY Reactor Recirculation pumps are running in slow speed.

ENSURE Main Turbine and Generator trip. (Reverse power 15 MCR 3 seconds time delay, 5 seconds time delay IF turbine has already tripped.).

a. VERIFY the Generator Output Breakers open.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 10 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES

b. VERIFY the Turbine Stop and Control Valves close.

WHEN reactor water level Can be restored AND maintained MCR 3 above 11.4 inches, THEN PERFORM the following to prevent Reactor water level from reaching Level 9 RFPT trip setpoint (58 in.):

IF Reactor pressure is dropping rapidly, THEN SELECT SPEED AUTO OR MANUAL on the running Reactor Feed Pump AND LOWER Reactor Feed Pump discharge pressure to MAINTAIN Reactor level below 58 inches.

TRANSFER Feedwater Control to Start-Up Level Control per SOI 04-1-01-N21-1. (Attachment VII of SOI 04-1-01-N21-1 May be used.)

ENSURE Scram Discharge Volume Vent AND Drain valves MCR 3 closed IOI 03-1-01-4 SCRAM Recovery INSERT all SRMs AND VERIFY response on SRM recorders. MCR 3 SWITCH IRM/APRM LVL recorders to IRM AND VERIFY MCR 3 neutron monitoring established on IRMs IF scram signal Can be cleared AND Reactor level AND MCR 3 pressure are stable, THEN RESET scram AND RETURN CRD System to normal as follows:

BYPASS Scram Instrument Volume High Level signal by PLACING CRD DISCH VOL HI TRIP BYP switches RPS Div 1, 2, 3, 4 to BYPASS.

RESET scram by PLACING SCRAM RESET handswitches RPS Div 1, 2, 3, 4 to RESET.

VERIFY all CRDs settle into Position 00.

IF any Control Rod is NOT at the 00 position, THEN PERFORM one notch insert to attempt to force the rod to settle into the 00 position.

WHEN CRD DISCH VOL WTR LVL HI TRIP annunciator is clear, THEN RETURN CRD DISCH VOL HI TRIP BYP switches to NORMAL.

VERIFY that the HCU scram accumulators have been recharged by OBSERVING the ACCUM FAULT indicating lights on 1H13-P680 are out.

THROTTLE G33-F102 to raise bottom head drain flow AND limit MCR 3 Bottom Head Drain Line Heatup/Cooldown to < 100°F/HR.

Bottom head drain flow greater than 250 gpm May be required.

IF Reactor water level is high, THEN REJECT water to Main Condenser per SOI 04-1-01-G33-1 to MAINTAIN level band.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 11 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601- MCR 3 19B to the TEST position.

VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms.

PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601- MCR 3 18B to the TEST position.

VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13-P601-19A-G3) Alarms.

ADJUST F033, RWCU SYS BLWDN F/D CONT VLV is ~ 10% MCR 3 Open.

OPEN OR CHECK OPEN the following valves: MCR 3 F028 RWCU BLWDN CTMT INBD ISOL 1H13-P680 F034, RWCU BLWDN CTMT OTBD ISOL1H13-P680 IF rejecting to main condenser, OPEN OR CHECK OPEN in the MCR 3 following order:

F046 RWCU BLWDN TO MN CNDSR 1H13-P680 F041 RWCU BLWDN TO MN CNDSR BYP 1H13-P680 F235 RWCU BLWDN TO MN CNDSR 1H13-P870-3C F234 RWCU BLWDN TO MN CNDSR 1H13-P870-9C IF desired, while rejecting during depressurized OR low pressure MCR 3 conditions, F031, RWCU BLWDN ORF BYP VLV May be Open to allow maximum flow Begin rejecting by SLOWLY OPENING F033, RWCU SYS MCR 3 BLWDN FLO CONT valve, AND IF necessary THROTTLING CLOSED F042 OBSERVE FI-R602, RWCU BLWDN FLO indicator on 1H13- MCR 3 P680 MONITOR reactor water level, blowdown flow AND area/room MCR 3 temperature indication while reject is in progress.

ENSURE Bypass valves are maintaining Reactor pressure MCR 3 IF proceeding to Cold Shutdown, THEN PERFORM Cooldown MCR 3 per Attachment II of IOI 03-1-01-3 concurrent with remaining steps of this Attachment.

DEPRESS the MHC START DVC LOWER pushbutton on MCR 3 1H13-P680-9C to reduce the MHC START DVC to Zero.

CONFIRM the following Bleeder Trip valves are Closed: MCR 3

a. N36-F013A, FW HTR 5A EXTR STM BTV

b. N36-F013B, FW HTR 5B EXTR STM BTV

c. N36-F012A, FW HTR 6A EXTR STM BTV

d. N36-F012B, FW HTR 6B EXTR STM BTV J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 12 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES

e. N11-F003A, MSR A 1ST STG RHT EXTR STM BTV

f. N11-F003B, MSR B 1ST STG RHT EXTR STM BTV ENSURE Seal Steam Pressure AND Reactor Feed Pump MCR 3 operation maintained by main steam CLOSE the following valves as soon as possible following N/A N/A Not Turbine trip at Gas Rack 1N44D001-N to isolate Hydrogen required to Pressure Regulators N44-PCV-F505 AND F506: be

a. N44-FA20 performed

b. N44-FA21 to Shut down and Cooldown the plant.

OBSERVE the following actions occur: MCR 3 Aux Field amps AND generator output voltage indicate 0. Generator Primary field breaker Will trip on a generator/transformer lockout Water Circ condition (including reverse power) AND the TVR feeder switch Pump can Will open IF a lockout was NOT initiated WHEN the Turbine be verified speed drops to ~1620 rpm. running TURB AUX OIL PMPS A, B OR C starts at about 1335 rpm. by computer AUX PW CIRC PUMP starts at about 815 rpm. (Locally) point in TURB SHAFT LIFT OIL PMP starts at about 510 rpm. the MCR.

TURB GEAR OIL VLVs N34-FE01/FE02 open at about 210 rpm.

THROTTLE P43-F053 to maintain Main Turbine Lube Oil temp N/A N/A Not between 90-119F. required to be performed to Shut down and Cooldown the plant.

WHEN fast speed trend recording is no longer necessary AND MCR 3 vessel level is greater than 11.4" AND vessel pressure is less than 1064.7 psig., THEN PERFORM the following:

DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON B21-R623A on 1H13-P601-20B.

DEPRESS the POST ACC MON HI SP RESET pushbutton for POST ACC MON B21-R623B on 1H13-P601-17B.

OPEN the Generator motor operated air break GEN DISC MCR 3 J5230.

PLACE Red Tag on the Control Room handswitch for J5230 in open position. (This step May be performed after step 9.30.3)

AFTER GEN DISC J5230 is opened, THEN PERFORM the MCR 3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 13 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES following:

IF tripped, THEN RESET the following Generator reverse power relays by PRESSING the relay reset rod upwards:

a. 432/G12 (1N41-M752)

b. 432/UT11 (1N41-M756)

AFTER Generator reverse power relays are reset, THEN RESET the following Generator Lockout relays, IF tripped:

a. 486-1/G12 (1N41-M769)

b. 486-2/G12 (1N41-M770)

c. 786-1/UT11 (1N41-M759)

d. 786-2/UT11 (1N41-M760)

AFTER all Generator Lockout relays are reset AND GEN UNIT MCR 3 TRIP annunciator clears on 1H13-P680-9A-A8, THEN OBTAIN Entergy Mississippi dispatchers permission AND PERFORM the following to close breakers J5228 AND J5232 from 1H13-P680 panel:

PLACE SYNC CONT BRKR J5228 switch to ON position.

CLOSE 500 KV BRKR J5228.

PLACE SYNC CONT BRKR J5228 switch to OFF position PLACE SYNC CONT BRKR J5232 switch to ON position.

CLOSE 500 KV BRKR J5232.

PLACE SYNC CONT BRKR J5232 switch to OFF position.

IF all Generator Lockout relays Will NOT reset, THEN MCR 3 PERFORM the following:

CONTACT Electrical Maintenance to investigate reason any other Generator relays other than reverse power May have tripped.

REQUEST Entergy Mississippi dispatcher to open disconnects to de-energize breaker(s) J5228 AND J5232.

DEPRESS EHC SP DEMAND LOWER AND REL pushbutton on MCR 3 1H13-P680-9C to reduce SP DEMAND indicator to 0 percent.

WAIT for SP LTD meter to decrease to 0 percent.

At each Main Transformer Control Cabinet (Phase A, Phase B, Outside at 3 Not AND Phase C), MN XFMRs Required VERIFY lead cooler group fans are OFF SECURE the following steam loads to limit plant cooldown:

• SJAE per SOI 04-1-01-N62-1 CLOSE Recombiner Drain Valves N64-F264 AND F265 (N64- 93 OG 3 Not F268 AND F269) Preheater Required A/B Rooms 1T109 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 14 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES 1T110 CLOSE N64-F007A(B) Preheater Inlet Drain using handswitch 113 Turb 3 Not on N64-P001. Area 1 Required 1T202 OPEN RECOMBINER AIR PURGE A(B) Manual Valve 1N64- 93 OG 3 Not F004A(B) Train A(B) Purge Air Sply Sol Byp for the Preheater Required corresponding recombiner train to ESTABLISH a purge flow of A/B Rooms approximately 60 scfm through the recombiner train.

1T109 1T110 CLOSE N62-F003A(B) CNDSR AIR TO 1 STG SJAE A(B) 133 Turb 3 Not locally at 1H22 P176 Area 1/4 Required OBSERVE that F003A(B) CNDSR AIR TO 1 STG SJAE A(B) 1T305, indicates Closed before continuing to the next step. 1T324 DEPRESS N62-F003A(B) SJAE A(B) 1ST STG SUCT VLV MCR 3 CLOSE pushbutton on 1H13-P680 [10C].

CHECK the indication on 1H13-P680 and the following valves MCR 3 Close:

SJAE A(B) 1ST STG STM INL VLV, N62-F024A(B)

SJAE ICNDSR DR VLV, N62-F011A(B)

SJAE A(B) 2ND STG SUCT VLV, N62-F006A(B)

SJAE A(B) MN STM SPLY VLV, N62-F001A(B)

SJAE A(B) EXH VLV, N62-F012A(B)

SJAE A(B) SEP DR VLV, N62-F002A(B)

REDUCE setpoint of N62-PIC-R010A(B) to zero 0 psi 113 Turb 3 Not Area 1 Required 1T202 ENSURE OPEN following handswitches on 1H22-P176: 133 Turb 3 Not N62-F004A the COND AIR TO 1 STG SJAE A Area 1/4 Required N62-F004B, the COND AIR TO 1 STG SJAE B 1T305, 1T324 ENSURE OPEN N62-F034 A, B, C, DISCH PIPE DRN VLV for 113 Turb A 3 Not draining discharge piping. MVP Area Required 1T218 WHEN discharge piping has drained, 113 Turb A 3 Not THEN CLOSE N62-F034 A, B, C, DISCH PIPE DRN VLV. MVP Area Required 1T218 OPEN N62-F014 MECH VAC PUMPS COM SUCT VLV, at 133 Turb 3 Not 1H22-P176. Area 1/4 Required J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 15 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES 1T305, 1T324 ENSURE proper mechanical vacuum pump oil level (>50%), 113 Turb A 3 Not THEN Prelube with manual oiler as follows: MVP Area Required ENGAGE manual oiler pump handle 1T218 ROTATE for a minimum of 60 seconds.

DEPRESS each plunger 5 times CHECK oil flow visible from each oil return line.

CLOSE P44-F348 A(B,C) MECH VAC PMP COOLER DRAIN. 113 Turb A 3 Not OPEN P44-F109 A(B,C) MECH VAC PMP PSW INL ISOL. MVP Area Required OPEN P44-F344 A(B,C) MECH VAC PMP PSW DISCH ISOL. 1T218 BLOW DOWN strainer as follows:

(1) OPEN P44-F316 A(B,C), MECH VAC PMPA(B)(C) STR DR.

(2) WHEN blowdown has been completed, THEN CLOSE P44-F316 A(B,C) MECH VAC PMPA(B)(C) STR DR.

START MECH VAC PMP A(B)(C) with START pushbutton on MCR 3 1H13 P680.

CHECK proper vacuum pump operation for each running pump 113 Turb A 3 Not by OBSERVING the following: MVP Area Required Cooling Water Inlet Valve 1P44-SV-F514A, B, OR C has opened 1T218 by MOMENTARILY OPENING drain valve 1P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR.

OBSERVING pressurized water flow, THEN CLOSE drain valve1P44-F348A, B, C MECH VAC PMP A(B)(C) CLR DR.

IF 1P44-SV-F514A, B, OR C did NOT open, THEN OPEN respective MECH VAC PMP A(B)(C) PSW SPLY BYP valve 1P44-F347A,B,C to provide cooling as needed for operation of Mechanical Vacuum Pump.

Suction Drain Valve SV-F507A, B OR C has Closed by OBSERVING no air suction flow.

Mechanical Vacuum Pump Inlet Valve F007A, B, OR C has Opened.

Proper oiler operation by OBSERVING oil flow from each oil return line.

Secure Seal Steam Generator per SOI 04-1-01-N33-1 PLACE Controller PK-R617 in MANUAL on 1H13-P878, AND MCR 3 CLOSE F506 as necessary to control reactor cooldown. The turbine Can be sealed with seal steam header pressure as low as 15 psig, PI-R622.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 16 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES Secure Reactor Feed Pump per SOI 04-1-01-N21-1 All areas previously addressed in securing 1st RFPT Offgas Preheater by placing controllers 1N64-R009A and 1N64- Turbine 3 Not R009B in manual and reducing output to 0 percent Building 93 Required Area 1 (1T113)

Main Steam Isolation valves AND/OR Main Steam Line MCR 3 SHUTDOWN a Condensate Booster Pump AND CLOSE MCR 3 respective discharge valve per SOI 04-1-01-N19-1, leaving one Condensate Booster Pump in service SHUTDOWN a Condensate Pump AND CLOSE respective MCR 3 discharge valve per SOI 04-1-01-N19-1, leaving one Condensate Pump in service CLOSE B21-F069 MCR 3 OPEN the following MSIV drain valves:

a. B21-F067A

b. B21-F067B

c. B21-F067C

d. B21-F067D OPEN the following valves: MCR 3

a. B21-F033

b. B21-F068 ISOLATE extraction steam to the HP Feedwater heaters as MCR 3 follows:

CLOSE the following valves:

a. N36-F010A EXTR STM SPLY TO FW HTR 5A

b. N36-F010B EXTR STM SPLY TO FW HTR 5B

c. N36-F011A EXTR STM SPLY TO FW HTR 6A

d. N36-F011B EXTR STM SPLY TO FW HTR 6B OBSERVE the following drain valves open: MCR 3 N36-F008A FW HTR 6A EXTR STM RO BYP DR VLV N36-F008B FW HTR 6B EXTR STM RO BYP DR VLV OPEN the following drain valves: MCR 3 OPEN HP Stop AND Control Valve Drain Valves by DEPRESSING each of the following MSCV UPSTRM DR VLV JOG OPEN pushbuttons:

a. N33-F078A J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 17 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES

b. N33-F078B

c. N33-F078C

d. N33-F078D OPEN Left Side Crossover piping drains by DEPRESSING each of the following XOVER PIPE LS DR VLV JOG OPEN pushbuttons:

a. N11-F043A (FR IST)

b. N11-F036A (FR 2ST)

c. N11-F044A (RE IST)

d. N11-F038A (RE 2ST)

OPEN Right Side Crossover piping drains by DEPRESSING each of the following XOVER PIPE RS DR VLV JOG OPEN pushbuttons:

a. N11-F044B (FR IST)

b. N11-F038B (FR 2ST)

c. N11-F043B (RE IST)

d. N11-F036B (RE 2ST)

OPEN N11-F015, MSCV A/B DNSTRM DR VLV.

OPEN the following MSR 2ND STG STM DR VLVS:

a. N11-F301

b. N11-F302 OPEN the following drain valves unless required closed to MCR 3 minimize cooldown:

OPEN Main Steam Line Drain Valves N11-F056, F055, F009, F011, F049, AND F050 by DEPRESSING MSL DR LINE ISOL VLVS OPEN pushbutton.

OPEN MSL Bypass Drain valves (N11-F002A, F002B, F002C, F002D, F010, F007, F052A, F052B, F057) using MSL DR VLVS DR LINE BYP VLV OPEN pushbutton.

DEPRESS Both NSSSS INBD ISOL RESET pushbutton (1H13- MCR 3 P601-18B) AND NSSSS OTBD ISOL RESET pushbutton (1H13-P601-19B) to reset logic AND re-energize RHR Logic lights on 1H13-P622 AND 1H13-P623 panels.

TRANSFER to startup level control IF NOT already in service MCR 3 TRANSFER the RFPT A(B) SP CONT to MAN. MCR 3 IF MSIVs are open with Main Condenser available, THEN MCR 3 INITIATE AND MAINTAIN cooldown at 90F/hr with one of the following methods:

CONTROL Reactor cooldown with Manual Bypass Jack on 1H13-P680-9C J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 18 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES At approximately 200 psig Reactor pressure, SHUTDOWN one RWCU Pump per SOI 04-1-01-G33-1, IF Both are running.

SLOWLY OPEN 1G33-F044, RWCU FLTR DMIN BYP VLV on MCR 3 Not 1H13-P680 while reducing F/D flow with flow controller 1G36- Required FC-R022A(B) on 1G36-P002. CTMT 185 RWCU Panel (1A509)

MAINTAIN a nearly constant system flow rate, (450-500 gpm MCR 3 Is recommended), as indicated on 1G33-FI-R609, RWCU INL FLO, on1H13-P680.

On 1G36-P002, OBSERVE that holding pump comes on WHEN CTMT 185 3 Not F/D flow is < 80%. RWCU Required Panel (1A509)

WHEN filter flow is < 20%, TURN Filter/Hold switch A(B) on CTMT 185 3 Not 1G36-P002 to HOLD position. RWCU Required OBSERVE the following valves fully Close: Panel

• G36-F001A(B) F/D Inlet (1A509)

• G36-F002A(B) F/D Inlet

• G36-F003A(B) F/D Outlet

• G36-F004A(B) F/D Outlet OBSERVE HOLD light on AND FILTER light out on 1G36-P002 CTMT 185 3 Not RWCU Required Panel (1A509)

PLACE the MANUAL/AUTO selector on controller 1G36-FC- CTMT 185 3 Not R022A (B) in MANUAL position with controller output at 0% RWCU Required output. Panel (1A509)

REPEAT Steps 4.6.2a AND 4.6.2b for second F/D. CTMT 185 3 Not RWCU Required Panel (1A509)

LOWER system flow rate to < 280 gpm by THROTTLING MCR 3 1G33F044 as indicated on 1G33FI-R609, RWCU INL FLO, on 1H13-P680.

TRIP one of the running RWCU pumps MCR 3 ESTABLISH 90 to 300 gpm flow as indicated on 1G33-FI-R609, MCR 3 RWCU INL FLO, on 1H13-P680 by THROTTLING the Bypass Valve 1G33F044 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 19 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES WHEN Reactor pressure is reduced to < 135 psig, THEN at approximately 40 psig, PLACE one loop of RHR System in SHUTDOWN COOLING mode per SOI 04-1-01-E12-2.

RACK OUT RHR A/B PMP Breaker, 152-1509/1606 Control 3 Required Bldg. 111 SWGR Rms 0C202, 0C215 SHUTDOWN RHR JOCKEY PUMP A/B on 1H13-P871. MCR 3 CLOSE F082A/B, RHR JCKY PMP SUCT ISOL VLV, on 1H13- MCR 3 P871.

CLOSE F064A/B, RHR MIN FLO TO SUPP POOL. MCR 3 CLOSE F004A/B, RHR PMP SUCT FM SUPP MCR 3 ENSURE OPEN F003A/B, RHR HX OUTL VLV MCR 3 ENSURE OPEN F048A/B, RHR HX A BYP VLV. MCR 3 CLOSE F047A/B, RHR HX INL VLV. MCR 3 CLOSE F428A/B, PRESSURE LOCK ISOL for F024 RHR A/B 3 Required Pump Rm Aux Bldg 93 (1A103/1A1 05)

CLOSE F438A/B, PRESSURE LOCK ISOL for F064 RHR A/B 3 Required Pump Rm Aux Bldg 93 (1A103/1A1 05)

SLOWLY OPEN F020, Manual Flush Valve. Aux Bldg 3 Required 119 RCIC Rm (1A204)

OPEN F006A, RHR PMP A SUCT FM SHUTDN CLG AND MCR 3 MONITOR RHR HR A STM press indicator for rise in pressure.

VENT Shutdown Cooling suction header as follows: Aux Bldg 3 Required (a) OPEN F323. 119 RCIC (b) OPEN F399. Rm (1A204)

(c) WHEN a solid stream of water is observed out of vent line, THEN CLOSE F399.

(d) CLOSE F323.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 20 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES OPEN F073A, RHR HX A OTBD VENT VLV. MCR 3 OPEN F074A, RHR HX A INBD VENT VLV. MCR 3 VENT RHR A Heat Exchanger A as follows: Aux. 139 3 Required (a) OPEN F400A, A RHR HX VENT. RHR A/B (b) OPEN F401A, A RHR HX VENT. Rm 1A303, 1A304/1A30 (c) WHEN water is observed from vent, THEN CLOSE F401A. 6, 1A307 (d) CLOSE F400A.

OPEN F064A/B. AFTER approximately one minute, THEN MCR 3 CLOSE F064A/B WHEN Conductivity as indicated on HX A/B OUT CNDCT, is as MCR 3 low as practical (Should be less than 2.0 µmhos/cm),

THEN CLOSE F073A/B, RHR HX A/B OTBD VENT VLV.

CLOSE F074A/B, RHR HX A/B INBD VENT VLV MCR 3 LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119 RCIC Rm (1A204)

CLOSE F048A/B MCR 3 OPEN F063A/B, Manual Flush Valve. RHR A/B 3 Required Pump Rm Aux Bldg 119 (1A203/1A2 05)

OPEN F073A/B, RHR HX A/B OTBD VENT VLV MCR 3 OPEN F074A/B, RHR HX A/B INBD VENT VLV MCR 3 WHEN Conductivity as indicated on HX A/B OUT CNDCT, is as MCR 3 low as practical (Should be less than 2.0 µmhos/cm),

THEN CLOSE F073A/B, RHR HX A/B OTBD VENT VLV.

CLOSE F074A/B, RHR HX A/B INBD VENT VLV MCR 3 LOCK CLOSED F063A/B, Manual Flush Valve. RHR A/B 3 Required Pump Rm Aux Bldg 93 (1A203/1A2 05)

OPEN F048A. MCR 3 OPEN F047A. MCR 3 J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 21 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES ENSURE OPEN F003A. MCR 3 ENSURE Shutdown Cooling Isolation Logic is reset by MCR 3 PRESSING NSSSS INBD ISOL RESET pushbutton AND NSSSS OTBD ISOL RESET pushbutton on 1H13-P601.

PLACE Standby Service Water A System in service to RHR A MCR 3 Heat Exchanger on 1H13-P870 as follows.

START SSW Pump A per SOI 04-1-01-P41-1.

OPEN P41-F014A, SSW INL TO RHR HX A.

ENSURE OPEN P41-F068A, SSW OUTL FM RHR HX A.

START RHR RM A FAN COIL UNIT.

ENSURE OPEN F010, SHUTDN CLG MAN SUCT VLV. MCR 3 ENSURE CLOSED F040, RHR TO RADWST OTBD SHUTOFF MCR 3 VLV.

ENSURE CLOSED F049, RHR TO RADWST INBD SHUTOFF MCR 3 VLV.

OPEN F020, Manual Flush Valve approximately 3 turns. Valve Aux Bldg 3 Required May be opened further IF required for level control. 119 RCIC Rm (1A204)

OPEN F008, RHR SHUTDN CLG OTBD SUCT VLV MCR 3 OPEN F009, RHR SHUTDN CLG INBD SUCT VLV as follows; MCR 3 ENSURE breaker 52-163137 is CLOSE position OPEN F009, RHR SHUTDN CLG INBD SUCT VLV MONITOR Reactor water level WHILE 1E12F009 AND 1E12F020 are OPEN.

PERFORM IMMEDIATELY the next step 4.1.2.b(14) IF a rise in Reactor water level is NOT desired.

LOCK CLOSED F020, Manual Flush Valve. Aux Bldg 3 Required 119 RCIC Rm (1A204)

NOTIFY Radwaste Operators to be prepared for Reactor water MCR 3 Required flush to Waste Surge tank.

Radwaste Building 118 Radwaste Control Room (0R241)

OPEN F203, RHR SYS FLUSH TO LIQ RADWST by the MCR 3 following handswitches to OPEN:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 22 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-3C)

F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C)

ENSURE CLOSED F070A/B, Manual RHR Drain Valve Aux Bldg 3 Required 93 Corridor (1A101)

OPEN F072A/B, RHR Drain Valve RHR A/B 3 Required Pump Rm Aux Bldg 93 (1A103/1A1 05)

SLOWLY OPEN F070A, RHR Drain Valve approximately one Aux Bldg 3 Required turn to start flow to Radwaste. 93 Corridor IF "RHR A DISCH PRESS ABNORMAL" annunciator alarms (1A101) while warming RHR A, THEN CLOSE F047A AND F048A to prevent draining of downstream piping.

THROTTLE F070A/B to warm RHR Pump A/B at less than Aux Bldg 3 Required 100°F/hr until RHR DISCH TO RADWST ON RHR TEMP 93 Corridor recorder is 200°F OR within 100°F of RX water temp, whichever (1A101) is less.

LOCK CLOSED F070A, RHR Drain Valve. Aux Bldg 3 Required 93 Corridor (1A101)

LOCK CLOSED F072A, RHR Drain Valve. RHR A/B 3 Required Pump Rm Aux Bldg 93 (1A103/1A1 05)

CLOSE F203, RHR SYS FLUSH TO LIQ RADWST by TAKING MCR 3 the following handswitches to CLOSE:

F203 SVA-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-3C)

F203 SVB-RHR SYS FLUSH TO LIQ RADWST (1H13-P870-8C)

RACK IN RHR A/B PMP Breaker, 152-1509/1606 Control 3 Required Bldg. 111 SWGR Rms 0C202, 0C215 NOTIFY Chemistry AND Radiation Protection that possibility of a MCR 3 crud burst Could occur due to starting of RHR pump in SDC mode J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 23 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES START OR ENSURE running RHR RM A FAN COIL UNIT on MCR 3 1H13-P870.

ENSURE CLOSED F064A, RHR A MIN FLO TO SUPP POOL. MCR 3 ENSURE RHR JOCKEY PUMP A is shutdown. MCR 3 ENSURE CLOSED F082A, RHR A JCKY PMP SUCT ISOL VLV. MCR 3 ENSURE CLOSED F004A, RHR A SUCT FM SUPP POOL. MCR 3 ENSURE OPEN the following valves: MCR 3 (a) F010 (Concurrent Verification Required)

(b) F008 (c) F009 as follows; (1) ENSURE breaker 52-163137 is CLOSE position (2) ENSURE OPEN F009, RHR SHUTDN CLG SUCT VLV (d) F006A (e) F047A (f) F048A CLOSE F003A, RHR HX A OUTL VLV . MCR 3 ENSURE CLOSED B21-F065A, FW INL SHUTOFF VLV. MCR 3 START RHR PMP A AND IMMEDIATELY FULLY OPEN one of MCR 3 the following valves:

(a) E12-F053A, RHR A SHUTDN CLNG RTN TO FW (b) E12-F037A, RHR A TO CTMT POOL (c) E12-F042A, RHR A INJ SHUTOFF VLV MONITOR RHR HX A differential temperature on RHR MCR 3 TEMPERATURE RECORDER as follows:

RHR HX A Point 1(inlet) - Point 5(outlet)

ESTABLISH a cool down rate of less than 90°F/hr, as follows: MCR 3 Slowly JOG OPEN F003A to allow flow through heat exchanger, AND MONITOR cooldown rate.

THROTTLE one of the following valves to maintain RHR pump flow ~8600 gpm AND RHR heat exchanger flow ~8200 gpm:

IF flow is through F053A, THEN THROTTLE F053A AS LONG AS flow through valve is maintained < 8550 gpm.

IF E12-F003A is closed while in SHUTDOWN COOLING, MCR 3 THEN MONITOR REACTOR COOLANT TEMPERATURE using the following indications:

REACTOR RECIRC LOOP A/B suction temperature (IF recirc pump(s) running)

RWCU REGENERATIVE HEAT EXCHANGER INLET J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 24 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES temperature (IF RWCU pump(s) are running.)

Point 5 of RHR TEMPERATURE RECORDER.

Installed thermocouple suspended above Reactor core.

WHEN F003A valve is full open AND additional cooling is MCR 3 required, THEN SLOWLY THROTTLE CLOSE F048A as needed to establish desired cooldown rate.

WHEN F048A valve is full closed, THEN, IF desired, THROTTLE MCR 3 F003A to MAINTAIN desired coolant temperature OR SDC flow while MAINTAINING 3000 gpm flow. F048A may be fully opened to reduce cooldown rate but CANNOT be left in a throttled position UNTIL F003A is full open.

SELECT "Shutdown Cooling-RHR A" OP GUIDE on PDS MCR 3 computer. The guide Should be left on-screen OR icond WHEN the respective shutdown cooling loop is in service until Reactor Coolant has been stabilized at desired temperature so that the guide Will warn operators IF Shutdown Cooling parameters are out of range LOG Reactor coolant temperature on Data Sheet I of 03-1-01-3 MCR 3 OR other applicable IOI. TAKE temperatures as required by 03-1-01-3 during cooldown AND CONTINUE to take readings once per hour WHEN temperature is stable.

LOG temperatures for SSW/RHR HX AND reactor coolant on log MCR 3 similar to Attachment I to ENSURE SSW temperature does NOT exceed design temperatures. (Ref. CR1997-0282)

IF SSW A auto start signal from RHR A pump running is defeated by Temporary Alteration, THEN START/STOP SSW A AND B fans as necessary to MAINTAIN SSW A Supply temp.

(E12-R601, pt. 12) between 50 AND 75 deg.

IF RPV level control via RWCU blowdown is unavailable, MCR 3 THEN RPV level control May be established by USING E12-F073A AND E12-F074A RHR heat exchanger vent to establish RPV level control, AND THROTTLE OPEN E12-F073A AND E12-F074A as required to establish AND maintain the desired RPV level.

MONITOR RPV level while reject is in progress.

IF desired to add water to Reactor with SDC in operation WHEN Aux Bldg 4, 5 Not in Modes 4 OR 5, THEN PERFORM the following: 119 RCIC Required THROTTLE OPEN, F020. Rm (1A204)

WHEN desired Reactor Vessel Level is reached, THEN LOCK CLOSED F020.

At approximately 120 psig, PERFORM the following:

TRANSFER RWCU to Pre-pump mode per SOI 04-1-01-G33-1.

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GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 25 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601- MCR 3 19B to the TEST position.

VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Alarms.

PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601- MCR 3 18B to the TEST position.

VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13-P601-19A-G3) Alarms.

SECURE RWCU blowdown flow per Section 5.1 of this MCR 3 instruction.

STOP running RWCU pump AND leave F044, RWCU FLTR MCR 3 DMIN BYP VLV THROTTLED SLIGHTLY OPEN.

CLOSE the following valves AND proceed to Step 4.4.2g without MCR 3 delay:

F250 RWCU SPLY TO RWCU HXS 1H13-P870-3C F251 RWCU SPLY TO RWCU HXS 1H13-P870-9C F252 RWCU HX RTN TO RWCU PMPS 1H13-P870-9C F253 RWCU HX RTN TO RWCU PMPS 1H13-P870-3C F255 RWCU FLTR/DMIN INL FM RWCU PMP 1H13-P870-5C OPEN OR CHECK OPEN the following valves: MCR 3 F004 PMP SUCT CTMT OTBD ISOL 1H13-P680 F001 PMP SUCT DRWL INBD ISOL 1H13-P680 F254 RWCU FLTR/DMIN INL FM RWCU HX 1H13-P870-5C F256 RWCU HX INL FM RWCU PMP 1H13-P870-5C CLOSE OR CHECK CLOSED F044, RWCU FLTR DMIN BYP MCR 3 VLV; And THEN RESTART one RWCU pump AND JOG OPEN F044 to establish flow greater than 90 gpm but less than 300 gpm.

START one RWCU the pump AND THROTTLE F044 to achieve MCR 3 a system flow greater than 90 gpm, But less than 300 gpm as indicated on FI-R609, RWCU INL FLO.

IF performing system warm-up. THEN MAINTAIN minimum flow, AVOIDING low flow trip.

IF desired, START a second pump as follows: MCR 3 START the pump AND THROTTLE F044 to maintain 300 - 500 gpm system flow as indicated on FI-R609, RWCU INL FLO, with Both Pumps running.

IF desired, ESTABLISH RWCU blowdown flow in accordance All areas with Section of this instruction previously addressed J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 26 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES for this evolution IF desired, PLACE F/Ds in service in accordance with Section All areas 4.5 of this instruction. previously addressed for this evolution PLACE NSSSS OTBD MOV TEST handswitch on 1H13-P601- MCR 3 19B to the NORM position.

VERIFY that "RX DIV 1 ISOL SYS OOSVC" annunciator (1H13-P601-19A-H3) Clears.

PLACE NSSSS INBD MOV TEST handswitch on 1H13-P601- MCR 3 18B to the NORM position.

VERIFY that "RX DIV 2 ISOL SYS OOSVC" annunciator (1H13-P601-19A-G3) Clears.

SHUTDOWN the running Condensate Booster Pump AND MCR 3 CLOSE respective discharge valve per SOI 04-1-01-N19-1.

IF scheduled, THEN PERFORM 06-OP-1B21-R-0010 (Att. I Not AND/OR II) WHEN reactor pressure is between 50 AND 100 required to psig be performed to Shut down and Cooldown the plant.

At approximately 60 psig Reactor pressure, PERFORM the MCR 3 following:

VERIFY that RCIC system isolates automatically.

IMMEDIATELY NOTIFY CAS, SAS, OR Security Island that RCIC is not available (non-functional).

COMPLETE shutdown of RCIC system per SOI 04-1-01-E51-1.

WHEN cooldown using Bypass Valves is no longer desired AND MCR 3 Shutdown Cooling is in service, THEN CLOSE the Bypass Valves as follows:

SET the TURB STM PRESS DEMAND setpoint approximately 100 psig above Reactor pressure using the PRESS REF RAISE OR LOWER pushbuttons on 1H13-P680-9C.

DEENERGIZE the Manual Bypass Valve Controller by depressing the MAN BYP CONT OFF pushbutton on 1H13-P680-9C.

IF MSIVs are open AND stroke time testing was NOT MCR 3 scheduled, THEN PERFORM the following:

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 27 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases IOI / SOI ACTIONS LOCATION MODE NOTES CLOSE the following Inboard MSIVs:

• B21-F022A

• B21-F022B

• B21-F022C

• B21-F022D WHEN Main Steam Line pressure downstream of MSIVs is near zero psig, THEN CLOSE the following Outboard MSIVs:

• B21-F028A

• B21-F028B

• B21-F028C

• B21-F028D CLOSE B21-F016 CLOSE B21-F019 NOTIFY Radiation Protection that the Reactor is to be vented to MCR 3 Drywell sump AND REQUEST Drywell survey after Head Vent realignment.

WHEN Reactor coolant temperature is less than 210F, THEN MCR 3 REALIGN Reactor Head Vents on 1H13-P601 as follows:

OPEN 1B21-F001, RPV OTBD VENT VLV.

PEN 1B21-F002, RPV INBD VENT VLV.

CLOSE 1B21-F005, RPV VENT TO MSL A.

Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2.

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 3 Page 28 of 28 Attachment 3 - Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Table A-3 & H-2 Results Table A-3 & H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Control Building 111 SWGR Rms (0C202, 0C215) 3 Auxiliary Building 93 RHR A Pump Room (1A103) 3 Auxiliary Building 93 RHR B Pump Room (1A105) 3 Auxiliary Building 93 Corridor (1A101) 3 Auxiliary Building 119 Corridor (1A201) 3 Auxiliary Building 119 RHR A Pump Room (1A203) 3 Auxiliary Building 119 RHR B Pump Room (1A205) 3 Auxiliary Building 119 RCIC Room (1A204) 3 Auxiliary Building 139 RHR A Room (1A303, 1A304) 3 Auxiliary Building 139 RHR B Room (1A306, 1A307) 3 Radwaste Building 118 Radwaste Control Room 3 (0R241)

J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 4 Page 1 of 2 GUIDELINES TO TERMINATE EMERGENCY PURPOSE: This Attachment provides general guidelines to be followed should changing plant conditions warrant termination of an emergency classification. The Emergency Director should use Attachment IV to terminate the emergency.

NOTE The Emergency Director discusses existing conditions with the Executive Director of Mississippi Emergency Management Agency (MEMA) or designee located in the MEMA Emergency Operations Center (EOC) and the Secretary - Louisiana Department of Environment Quality (LDEQ) located in the Louisiana State EOC prior to terminating an emergency that reached a Site Area Emergency (SAE) or General Emergency (GE) classification.

I. Termination Guidelines A. General

1. Conditions which caused the event have been terminated.

2. Circumstances which have arisen from the event are under control and the results of any and all pertinent data are evaluated.

3. All probability of recurrence of an event is removed, isolated or under control.

B. Examples CATEGORY l TERMINATION GUIDELINES l

Fires l Removal/separation of any element of fire l triangle. Fire under control/not spreading.

l l

Spill l Tanks, pipes, valves, any other problem sources l are empty, isolated, and out of service.

l l

Airborne l Source identified and isolated and/or contained.

l Area controlled.

l l

Explosion l Existing and potential hazards removed, destroyed l and/or isolated.

l l

Abnormal l Liquid discharge is terminated, sampling is completed, Effluent l and statistics verified. Public exposure to Offsite l radioactive material is reduced or eliminated.

l l Airborne - Source identified and analysis complete.

l Release is terminated and its cause is under control.

l All Onsite and Offsite monitoring data is evaluated.

l Public exposure to Offsite radioactive material is l reduced or eliminated.

l J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 4 Page 2 of 2 GUIDELINES TO TERMINATE EMERGENCY B. Examples (Cont.)

CATEGORY l TERMINATION GUIDELINES l

Control Room l Plant in normal emergency shutdown from remote Evacuation l stations. Cause of evacuation identified and under l control. No radiological conditions exist which l cause the Control Room to become uninhabitable.

l l

Plant Shutdown l Unit is shut down by normal or emergency means.

Functions (not l Unit is in cold shutdown and there is no potential available or l for uncontrolled criticality.

failed) l l

l Fuel Handling l Fuel elements, segments, pellets not in a critical Accident - New l configuration. Airborne activity has been evaluated or Spent Fuel l and accountability of components complete.

Damage, l Channeled or l Unchanneled l l

l Water Loss - l Source of water loss is defined. Ability to restore or LOCA Abnormal l maintain water level adequate for proper cooling.

Primary l Coolant Leak l l

l Earthquake or l The plant has been returned to a safe condition.

Other Natural l Threat of aftershock has passed and any damage Disaster l has been evaluated as to risk, if any.

l l

Security Threatl Threat to site is terminated. Probability of l recurrence has been removed, with the concurrence l of Security Supervisor and State, Local and l Federal Officials.

l J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 5 Page 1 of 1 Event Termination Checklist INSTRUCTION This checklist may be used by the Emergency Director to evaluate a decision to terminate an existing emergency condition. All criteria shall be met.

This checklist, completed and signed by the Emergency Director, is a prerequisite for initiation of the Recovery Organization.

CRITERIA CRITERIA MET (Initials)

1. The plant is in a stable configuration _________

with adequate core cooling.

2. In-plant radiation levels are stable _________

or decreasing with time.

3. The release of radioactive material to the _________

Environment is controlled and there is no significant potential for additional uncontrolled release.

4. All safety systems necessary to maintain the __________

plant in a stable configuration are operable.

5. Fires are extinguished; flooding conditions __________

and any other site damage is under control.

6. All vital areas requiring occupancy are __________

habitable.

7. Site security control is established. __________

_________________________________Date ________________ Time __________

Emergency Director J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-1 Revision 131 Attachment 6 Page 1 of 1 Rapidly Progressing Severe Accident Determination General Emergency Declared Will this be No NOT a Rapidly the first PAR for Progressing Severe the event? Accident Yes Is there No a LOSS of the containment fission product barrier as defined by the EALs?

No Yes A radiologica No release expected to start Containment Higli within ~1 hour with site boundary Rad Monitor>=

Dose projections >1,000 mR TEDE 10,000 R/hr?

or 5,000 mR CDE Thyroid?

Yes Yes A Rapidly Progressing Severe Accident is in progress ~------------___J J:\ADM_SRVS/TECH_PUB/REVISION/10/10S011.DOC

Attachment 4 Page 1 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 132 SECTION 1- PROCEDURE INFORMATION Procedure

Title:

Activation of the Emergency Plan PRGGN/CR No.: PRGGN 2020-00092 Database File ID: - -N/A TCN Number(s): _N_/A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Level of Use Changed Procedure Title Changed Procedure Number Changed Revision Change Details/Justifications:

(Pages 3 and 4, Revision Change Details - Continued may be used, if needed.)

See continuation pages.

SECTION 2 - WORKFLOW PUP: Total Rewrite for Human Factoring P-FIN: Maintains Existing Format New Procedure Minor Revision

~ Major Revision SECTION 3 - REVIEWS Review Type: Print Name Date Procedure Writer: D.M. Coulter 1-29-20 Technical Review: J.J. Lewis Pw 2-18-20 CDR Review:

• Wes Marshall 2-7-20 Validation Review:

• Tre Reeves 2-6-20 2 nd Review:

• N/A

• Reviews not required may be marked N/A 01-S-02-9 R010

Attachment 4 Page 2 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 132 SECTION 4 - REVIEW/EVALUATION Reviews: (MARK appropriate box, Print Name, Sign, and Date - Signature Required)

PAD Not Required Process Applicability Excluded IS/IST Implementation TCN Incorporation PER EN-Ll-100 or Other P r o c e s f J N u m ~

01-S-02-9 IX! PAD Required D.M. Coulter

• L*{7*tO PER EN-Ll-100 Print Name Signature Date X Additional reviews List additional reviews performed and attach copy of reviews to PRGGN:

were performed and listed in section IV of 50.54Q the PAD form.

SECTION 5 - SUPERVISOR REVIEWS Change Management Plan Required: IZI N/A Yes PRGGN CA initiated: IZI No Yes PRGGN CA No.:

Remarks:

Change Management Plan is not Required. All users of the procedure attended Class Room Training GLP-EP-EALRev6, which reviewed the revised Emergency Action Levels OSRC Review Required: IZI No Yes Michael L. Lewis Print Name

/41-~> Signature

z. Z.8

• z.o Date SECTION 6 - OSRC REVIEW IZI N/A Section APPROVED REJECTED Meeting No.:

N/A Print Name Signature Date SECTION 7 - PROCEDURE OWNER or DESIGNEE APPROVAL Michael L Lewis Print Name

,t1// /,~J

{ Signature

-;. . 2-S . z_o Date 01-S-02-9 R010

Attachment 4 Page 3 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 132 Revision Change Details - Continued This page is submitted into Records only if used.

Page: - - 1 - of - - 2 -

The following are changes to the 10-S-01-1 Change# Page Description of Change 1 Flow Charts Changes made for ease of use. Added background color of referred section to notes.

1.a Cold S/D Changed Cold S/D category background color from salmon Category to light blue 1.b CG UE Block Added reference under C6 UE block and S8 UE Block with S8 UE Block yellow background: "Refer to Category H for potential UE due to a Hazardous Event" 1.c S4 Alert Block Added note with green background under S4 Alert block:

"Refer to Table F-1 for potential escalation due to degraded fuel clad barrier."

1.d SS Alert Block Added note with green background under S5 Alert block:

"Refer to Table F-1 for potential escalation due to degraded RCS barrier."

1.e H2 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a seismic event" 1.f H3 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a hazardous event" 1.g H4 Cold Alert Revised note and added light blue background color to read:

Block "Refer to CA6.1 for potential escalation due to a fire" 1.h H2 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a seismic event" 1.i H3 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a hazardous event."

1.j H4 Hot Alert Revised note and added magenta background color to read:

Block "Refer to SA8.1 for potential escalation due to a fire."

2 Flow Charts Added the computer monitoring points for release pathways in Table A-1 01-S-02-9 R010

Attachment 4 Page 4 of 4 Procedure Traveler Procedure Number: 10-S-01-1 NEW Rev Number: 132 Revision Change Details - Continued This page is submitted into Records only if used.

Page: - - 2 - of - -

3 Flow Charts Added page number reference for EAL basis document for each EAL 4 Flow Charts Added procedure number and revision date s Flow Charts Added controlled document number box.

6 Flow Charts Changed EAL CAl.1 RPV water level (Level 2) from <-42 in.

to <-41.6 in.

7 Attachment 2 Changed EAL CAl.1 RPV water level (Level 2) from <-42 in.

CAl.1 Pages 80 to <-41.6 in. Changed reference to GGNS TRM Table and 81 TR3.3.4.2-1 8 Attachment 2 Corrected second reference from 1 to 2.

CNB2 Page 161 9 Attachment 2 Shaded the EAL boxes of all EALs for ease of locating the EALs 10 Attachment 2 Removed numerical sub-contents. Removed "7 .0 page 1 Table of Attachments" from TOC and replaced with EAL Listing and Contents page numbers.

11 Attachment 3 Corrected degree symbol for the Main Turbine Lube Oil Page 12 temperature range.

12 Procedure Page Revised Attachment numbering. Attachment 1 is now the 4 and flow charts with the remaining attachments increased by Attachment one. Changed the attachment numbers from Roman Pages numerals to Arabic numbers.

01-S-02-9 R010