RA-20-0009, Emergency Plan, Revision 20-01
| ML20015A466 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/15/2020 |
| From: | Simril T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-20-0009 | |
| Download: ML20015A466 (388) | |
Text
(_~ DUKE Tom Simril ENERGY Vice President Catawba Nuclear Station Duke Energy CN01VP 14800 Concord Road York, SC 29745 o 803 701 .3340 f 803 701 .3221 tom .simril@duke-energy.com RA-20-0009 10CFR 50.54(q)
January 15, 2020 U.S. Nuclear Regulatory Commission Attention : Document Control Desk Washington, DC 20555-0001
Subject:
Duke Energy Carolinas , LLC Catawba Nuclear Station Units 1 and 2 Docket Nos. 50-413 and 50-414 Emergency Plan , Revision 20-01 Enclosed for NRC Staff use is Revision 20-01 to the Cataw ba Nuclear Station Emergency Plan. This revision is effective as of December 16, 2019.
A summary of analysis has been provided for all changes. This summary is included in the attached 50.54(q) evaluation for the revision.
This revision is being submitted in accordance with 10CFR 50.54(q) and does not constitute a reduction in the effectiveness of the Emergency Plan for Catawba Nuclear Station. The 10CFR 50.54(q) Evaluation for Revision 20-01 to the Emergency Plan is provided as Attachment 1.
If there are any questions, please contact Staci White at 803-701-5191.
Sincerely, Tom Simril Vice President, Catawba Nuclear Station Attachments : 1. Emergency Plan 10CFR50.54 (q) Evaluation(s) and screen (s)
- 2. Plan update instructions
- 3. Emergency Plan, Revision 20-01 www duke-energy.com
Catawba Nuclear Station E-Plan Revision 20-01 RA-20-0009 January 15, 2020 Page 2 xc (w/attachments):
Laura Dudes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 xc (w/o attachments):
Michael Mahoney NRC Project Manager (CNS)
U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop O8B1A 11555 Rockville Pike Rockville, MD 20852-2738 Joseph D. Austin NRC Senior Resident Inspector (CNS)
Catawba Nuclear Station Attachment 1 Emergency Plan Revision 20-01 10CFR 50.54(q) Evaluation(s) and screen(s)
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02288674 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02288674 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 11/21/2019 Pri Resp Fac :
Status : COMPLETE Reschedule :0 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 18255 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes I have reviewed and approved the changes to EPA D rev 148 and the 50.54Q screening.
Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 08/26/2019 DJC1758 INPROG 11/18/2019 I44004 11/21/2019 11/18/2019 I44004 NTFY/ASG 11/18/2019 I44004 ACC/ASG 11/18/2019 I44004 AWAIT/C 11/21/2019 MRCOYL1 ACC/ASG 11/21/2019 I44004 AWAIT/C 11/21/2019 MRCOYL1 ACC/ASG 11/21/2019 I44004 AWAIT/C 12/16/2019 MEHARE COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By editorial box needs to be checked "no" 20191121 MRCOYL1 editorial box needs to be checked "no" 20191121 MRCOYL1 Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02288674 Duke Energy ASSIGNMENT NBR - 01 Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MRCOYL1 A 11/18/2019 1430 RETURNED 11/21/2019 0746 COYLE MEHARE A 11/21/2019 1322 APPROVED 12/16/2019 1438 HARE MRCOYL1 A 11/21/2019 1212 APPROVED 11/21/2019 1322 COYLE MRCOYL1 A 11/21/2019 1134 RETURNED 11/21/2019 1203 COYLE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :2
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 1 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02288674 CNS CR3 HNP MNS 5AD #: 02297412 ONS RNP GO Document and Revision I EPA D, rev 148 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
Changes to the Catawba Nuclear Station EALs based on the clarifications provided by Emergency Preparedness Frequently Asked Questions (EPFAQs) 2015-013 (EAL HG1.1) and 2016-002 (EALs CA6.1 and SA9.1.) No other changes to the EALs have been made in this revision.
Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
ML19058A632, Subject CATAWBA NUCLEAR STATION, UNITS 1 AND 2; MCGUIRE NUCLEAR STATION, UNITS 1 AND 2; OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3; BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1; AND H. B.
ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENTS TO REVISE EMERGENCY ACTION LEVEL SCHEMES TO INCORPORATE CLARIFICATIONS PROVIDED BY EMERGENCY PREPAREDNESS FREQUENTLY ASKED QUESTIONS 2015-013, 2015-014, AND 2016-002 (EPID L-2018-LLA-0174), dated July 1, 2019, give approval for the changes to the Catawba Nuclear Station EALs based on the clarifications provided by Emergency Preparedness Frequently Asked Questions (EPFAQs) 2015-013 (EAL HG1.1) and 2016-002 (EALs CA6.1 and SA9.1).
Printed : 12/17/2019 Page : 3 Back
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 2 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Catawba Nuclear Station, Unit 1, Amendment 303 to Renewed License NPF-35 Catawba Nuclear Station, Unit 2, Amendment 299 to Renewed License NPF-52 Bounding document attached (optional) I Part III. Editorial Change No or Is this activity an editorial or typographical change only, such as formatting, Yes I Partially I paragraph numbering, spelling, or punctuation that does not change intent? 10 CFR 50.54(q) Continue to Effectiveness Attachment 4, Evaluation is not Part IV and Justification:
required. Enter address non justification and editorial complete changes Attachment 4, Part V.
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that on shift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of on shift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures Printed : 12/17/2019 Page : 4
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 3 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3)
Printed : 12/17/2019 Page : 5
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 4 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b) (10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b) (11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b) (12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b) (13) Recovery Planning and Post-Accident Operations 13a Plans for recovery and reentry are developed.
14 10 CFR 50.47(b) (14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 5 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
15 10 CFR 50.47(b) (15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b) (16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete , 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle See CAS See CAS Approver (Manager, Nuclear Support Services) Approver Signature: Date:
Name (Print): See CAS See CAS Mandy Hare Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
Printed : 12/17/2019 Page : 7
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 6 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
QA RECORD Printed : 12/17/2019 Page : 8
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, ACTION REQUEST - 02303163 Duke Energy ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02303163 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 11/21/2019 Pri Resp Fac :
Status : COMPLETE Reschedule :0 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 18255 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes I have reviewed and approve 50.54Q for superseding EPA D.
Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 11/18/2019 I44004 INPROG 11/18/2019 I44004 11/21/2019 11/18/2019 I44004 NTFY/ASG 11/18/2019 I44004 ACC/ASG 11/18/2019 I44004 AWAIT/C 11/21/2019 MRCOYL1 ACC/ASG 11/21/2019 I44004 AWAIT/C 12/16/2019 MEHARE COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By sect D is rev 147 and 30 day submittal?? 20191121 MRCOYL1 Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02303163 Duke Energy ASSIGNMENT NBR - 01 Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MRCOYL1 A 11/18/2019 1626 RETURNED 11/21/2019 1011 COYLE MEHARE A 11/21/2019 1219 APPROVED 12/16/2019 1445 HARE MRCOYL1 A 11/21/2019 1138 APPROVED 11/21/2019 1219 COYLE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :2
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 1 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02303163 CNS CR3 HNP MNS 5AD #: 02297412 ONS RNP GO Document and Revision I EPA D, Emergency Action Levels, Rev 149 (SUP)
Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
EPA D, Rev 148, will be superseded by CSD-EP-CNS-0101-01, Rev 00, in its entirety. No revisions or deletions were made to EPA D when it was converted to CSD-EP-CNS-0101-01, Rev 00.
Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional)
Part III. Editorial Change No or Is this activity an editorial or typographical change only, such as formatting, Yes I Partially paragraph numbering, spelling, or punctuation that does not change intent? 10 CFR 50.54(q) Continue to Effectiveness Attachment 4, Evaluation is not Part IV and Justification:
required. Enter address non justification and editorial complete changes Attachment 4, Part V.
Printed : 12/17/2019 Page : 3 Back
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 2 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that on shift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of on shift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures 5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3)
Printed : 12/17/2019 Page : 4
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 3 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b) (10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b) (11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b) (12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b) (13) Recovery Planning and Post-Accident Operations 13a Plans for recovery and reentry are developed.
14 10 CFR 50.47(b) (14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
Printed : 12/17/2019 Page : 5
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 4 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
15 10 CFR 50.47(b) (15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b) (16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
EPA D, Rev 148, will be superseded by CSD-EP-CNS-0101-01, Rev 00, in its entirety. No revisions or deletions were made to EPA D when it was converted to CSD-EP-CNS-0101-01, Rev 00. Planning Standard 4a is not affected by the change in the title of this document.
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete , 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle See CAS See CAS Approver (Manager, Nuclear Support Services) Approver Signature: Date:
Name (Print): See CAS See CAS Mandy Hare Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 5 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 12/17/2019 Page : 7
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02310566 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02310566 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 01/23/2020 Pri Resp Fac :
Status : COMPLETE Reschedule :0 Pri Resp Group :
Assigned To : MATTHEW L NELSON Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 44278 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 01/13/2020 I18938 INPROG 01/13/2020 I18938 01/23/2020 01/13/2020 I18938 ACC/ASG 01/14/2020 I18938 AWAIT/C 01/14/2020 DATHOMP COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By
- No Routing Comments Found ***
Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Action Request Assignment Completion Approval Printed : 01/15/2020 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02310566 Duke Energy ASSIGNMENT NBR - 01 Route List : 001 Route List Initiator : I18938 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MEHARE A 01/14/2020 1638 APPROVED 01/14/2020 1724 HARE E39701 A 01/14/2020 1609 APPROVED 01/14/2020 1638 WHITE DATHOMP A 01/14/2020 1724 APPROVED 01/14/2020 2206 THOMPSON Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 01/15/2020 Page :2
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 1 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: Revised EREG 02310566 02303141 CNS This is a revised EREG to complete actions identified in AFI from CR3 2020 CNS NRC EP inspection readiness assessment. Additions will HNP be identified with bold italics and deletions will be identified with a strikethrough.
MNS 5AD #: 02297414 ONS RNP GO Document and Revision I CSD-EP-CNS-0101-01, EAL Technical Basis Document, Rev 0 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
This is a new document converted from EPA D, Emergency Action Levels, Rev 148.
Changes to EPA D, Rev 148 are as follows:
Page 3: Section 1.0 "Purpose"
- 1st paragraph, 3rd sentence - Removed "implementation of RP/0/A/5000/001".
Page 6: Section 2.4 EAL Organization nd
- Last Paragraph, 2 sentence - Added space after NEI 99-01 From: The CNS EAL categories are aligned to and represent the NEI 99-01Recognition Categories.
To: The CNS EAL categories are aligned to and represent the NEI 99-01 Recognition Categories.
Page 9: Section 2.6 "Operating Mode Applicability"
- Corrected the duplication of the number 3 (listed twice)
From To 2 .6 Operating Mode Appl icabi lity (ref. 4.1.7) 2.6 Operating Mode Applicability (ref. 4.1 .7) 1 Power Operation 1 Power Operation K.. "'.. 0.99 and reactor thennal power > 5% Ke*"'._ 0.99 and reactor thermal power > 5%
2 Startup 2 Startup K., "'._ 0.99 and reactor thermal power::,_ 5% Ke*"'._ 0.99 and reactor tl)_ermal power ::,_ 5%
3 Hot standby 3 Hot Standbt K., < 0.99 and average coolant temperature"'.. 350°1' Ke*< 0.99 and average coolant temperature"'._ 350°F 3 Hot Shutdown 4 Hot Shutdown K.. < 0.99 and average coolant temperature 350°F > T,.. > 200 °F K,, < 0.99 and average coolant temperature 350°F > T.,0 > 200 ' F 4 Cold Shutdown 5 Cold Shutdow n K.* < 0.99 and average coolant temperature::,_ 200"F K,, < 0.99 and average coolant temperature::,_ 200°F 5 Refueling Ito 6 Refueling One or more reactor vessel head dosure bolts are less than fully tensioned One or more reactor vessel head closure bolts are less than fu lly tensioned D Defueled D Defueled Reactor vessel contains no irradiated fuel Reactor vessel contains no irradiated fuel IPrinted I
- 01/15/2020 Page : 3 I Back
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 2 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Page 15: "Implementing"
- Replaced "RP/0/A/5000/001 Classification of Emergency" with "AD-EP-ALL-0101 Emergency Classification".
Page 16: Section 5.1 "Definitions"
- "Confinement Boundary" - Changed MAGNASTAR" to "MAGNASTOR".
Page 40: "CNS Basis Reference(s)"
- Removed "HP/0/B/1009/026 On-Shift Offsite Dose Assessment" Page 41: "Basis"
- Added HP/0/B/1009/014, Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material (ref. 3) and to the first paragraph.
Page 42: "CNS Basis Reference(s)"
- Added "HP/0/B/1009/014, Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material" as reference 3.
Page 43: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2),. Deleted (rev. 1).
Page 44: "CNS Basis Reference(s)"
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
Page 49: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2) of CNS.
Page 50: "CNS Basis Reference(s)"
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
Page 55: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
Page 56: "CNS Basis Reference(s)"
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
th Page 83: Basis, Last sentence of 7 paragraph
- Removed Printedextra: space after NCS.
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
Page 88: Basis, Last sentence of 2nd paragraph
- Added space between theNCS to read the NCS.
Page 88: Basis, First sentence of 2nd paragraph
- Added space between limitand to read limit and.
Page 108: "Basis"
- Added the following paragraphs:
"An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under CA6, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under CA6, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent."
Page 160: EAL,
- Added space between capabilityto to read capability to.
Page 176: "Basis"
- Added "Failure to isolate the leak within 15 minutes, or if known that leak cannot be isolated within 15 minutes from the start of the leak, requires immediate classification" in second to last paragraph.
Page 179: Basis, First sentence of first paragraph
- Removed additional space after trip on third line.
Page 179: Basis, First sentence of second paragraph
- Removed additional space after trip.
Page 181: Basis, First sentence of fourth paragraph rd
- Removed additional space after trip on first line and after trip on 3 line.
Page 181: Basis, First sentence of fifth paragraph
- Removed additional space after trip on third line.
Page 184: Basis, First sentence of fifth paragraph
- Removed additional space after trip on third line.
Page 193: EAL,
- Added space after between (Note 11,12) to read (Note 11, 12).
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
Page 195: Basis
- Added the following paragraphs:
"An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under SA9, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA9 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under SA9, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent."
Page 198: "EAL"
- In "Notification of Unusual Event" box, deleted "Vertical Storage Cask (VSC) > any of the following:" and "bullets" and replaced with "spent fuel cask> any Table E-1 dose limit."
- New E-1 dose limit table added.
nd Page 206, 2 Paragraph, E, change from Emergency Coordinator Judgment to Emergency Coordinator Jugement.
Page 223: "Basis" Added "If EOPs direct operators to open the Pressurizer pressure relief valves to implement a core cooling strategy (i.e., a feed and bleed cooldown), then there will exist a reactor coolant flow path from the RCS, past the pressurizer safety and relief valves and into the containment that operators cannot isolate without compromising the effectiveness of the strategy (i.e., for the strategy to be effective, the valves must be kept in the open position); therefore, the flow through the pressure relief line is UNISOLABLE. In this case, the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been eliminated and thus this condition constitutes a loss of the RCS barrier."
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ATTACHMENT 4 Page 5 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No I 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional) I I
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part III. Editorial Change No or Yes Is this activity an editorial or typographical change only, such as formatting, Partially paragraph numbering, spelling, or punctuation that does not change intent?
Justification:
Editorial Justification: AD-EP-ALL-0602, step 3.0.6.g, "Correct references or annotations that are no longer applicable"
- Page 3: Section 1.0 "Purpose"
- 1st paragraph, 3rd sentence - Removed "implementation of RP/0/A/5000/001".
- Page 15: "Implementing"
- Replaced "RP/0/A/5000/001 Classification of Emergency" with "AD-EP-ALL-0101 Emergency Classification".
- Page 40: "CNS Basis Reference(s)"
- Removed "HP/0/B/1009/026 On-Shift Offsite Dose Assessment" This procedure has been deleted.
- Page 43: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2). Deleted (rev. 1).
- Page 44: "CNS Basis Reference(s)"
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
- Page 49: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
- Page 50: "CNS Basis Reference(s)"
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
- Page 55: "Basis"
- Replaced HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS, with AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2).
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
- Page 56: "CNS Basis Reference(s)" 10 CFR 50.54(q) Continue to
- Removed HP/0/B/1009/004, Environmental Monitoring for Emergency Effectiveness Attachment 4, Conditions Within the Ten Mile Radius of CNS, and added AD-EP-ALL- Evaluation is not Part IV and 0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP- required. Enter address non CNS-0203, CNS Site Specific Field Monitoring (ref. 2). justification and editorial complete changes Attachment 4, Part V.
AD-EP-ALL-0602, step 3.0.6.a, "Correct typographical errors"
- Page 9: Section 2.6 "Operating Mode Applicability"
- Corrected the duplication of the number 3 (listed twice)
- Page 16: Section 5.1 "Definitions"
- "Confinement Boundary" - Changed MAGNASTAR" to "MAGNASTOR".
- Page 6: Section 2.4 EAL Organization Last Paragraph, 2nd sentence - Added space after NEI 99-01 From: The CNS EAL categories are aligned to and represent the NEI 99-01Recognition Categories.
To: The CNS EAL categories are aligned to and represent the NEI 99-01 Recognition Categories.
- Page 83: Basis, Last sentence of 7th paragraph
- Removed extra space after NCS.
- Page 88: Basis, Last sentence of 2nd paragraph
- Added space between theNCS to read the NCS.
- Page 88: Basis, First sentence of 2nd paragraph
- Added space between limitand to read limit and.
- Page 160: EAL,
- Added space between capabilityto to read capability to.
- Page 179: Basis, First sentence of first paragraph
- Removed additional space after trip on third line.
- Page 179: Basis, First sentence of second paragraph
- Removed additional space after trip.
- Page 181: Basis, First sentence of fourth paragraph
- Removed additional space after trip on first line and after trip on 3rd line.
- Page 181: Basis, First sentence of fifth paragraph
- Removed additional space after trip on third line.
- IPrinted Page 184:: Basis, 01/15/2020First sentence of fifth paragraph Page : 9 I
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 8 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
- Removed additional space after trip on third line.
- Page 193: EAL,
- Added space after between (Note 11,12) to read (Note 11, 12).
- Page 206, 2nd Paragraph, E, change from Emergency Coordinator Judgment to Emergency Coordinator Jugement.
AD-EP-ALL-0602, step 3.0.6.e, "Correct references..."
- Page 41: "Basis"
- Added HP/0/B/1009/014, Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material (ref. 3) to the first paragraph.
- Page 42: "CNS Basis Reference(s)"
- Added "HP/0/B/1009/014, Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material" as reference 3.
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that on shift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of on shift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures Printed : 01/15/2020 Page : 10
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3)
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EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 10 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b) (10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b) (11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b) (12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b) (13) Recovery Planning and Post-Accident Operations 13a Plans for recovery and reentry are developed.
14 10 CFR 50.47(b) (14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
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EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 11 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
15 10 CFR 50.47(b) (15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b) (16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below. -
Justification:
The changes remaining that are not editorial simply add clarification to existing EALs and do not affect planning standard 4a. This clarifying information enhances the user's understanding of the EAL, but is not technical in nature and is not required to correctly declare an EAL.
Page 176: "Basis"
- Added "Failure to isolate the leak within 15 minutes, or if known that leak cannot be isolated within 15 minutes from the start of the leak, requires immediate classification" in second to last paragraph.
The above change further clarifies Note 1 associated with EAL SU5.1. Note 1 states The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. The statement Failure to isolate the leak within 15 minutes, or if known that the leak cannot be isolated within 15 minutes from the start of the leak, requires immediate classification adds additional clarification.
Adding this statement to the basis of this EAL does not change the EAL scheme, the intent of the EAL or delay the classification of the EAL. No further evaluation is needed.
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
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ATTACHMENT 4 Page 12 of 12
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White Matthew Nelson See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle Eric White See CAS See CAS Approver (Manager, Nuclear Support Services) Approver Signature: Date:
Name (Print): See CAS See CAS Mandy Hare Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A David Thompson N/A See CAS N/A See CAS If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 01/15/2020 Page : 14
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, ACTION REQUEST - 02310566 Duke Energy ASSIGNMENT NBR - 06 Action Request Assignment Details AR NUMBER : 02310566 ASSIGNMENT NUMBER : 06 Type : EP02 Due Date : 01/23/2020 Pri Resp Fac :
Status : COMPLETE Reschedule :0 Pri Resp Group :
Assigned To : MATTHEW L NELSON Sec Resp Fac :
Subject : 50.54(Q) Evaluation Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 44278 Discipline :
Est Manhrs :0 Est Comp Date :
Description Complete 50.54(Q) Evaluation in accordance with AD-EP-ALL-06 02 Action Request Assignment Completion Notes Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 01/13/2020 I18938 INPROG 01/23/2020 01/13/2020 I18938 ACC/ASG 01/14/2020 I18938 AWAIT/C 01/14/2020 DATHOMP COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By
- No Routing Comments Found ***
Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Action Request Assignment Completion Approval Printed : 01/15/2020 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02310566 Duke Energy ASSIGNMENT NBR - 06 Route List : 001 Route List Initiator : I18938 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name E39701 A 01/14/2020 1612 APPROVED 01/14/2020 1641 WHITE MEHARE A 01/14/2020 1641 APPROVED 01/14/2020 1725 HARE DATHOMP A 01/14/2020 1725 APPROVED 01/14/2020 2202 THOMPSON Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 01/15/2020 Page :2
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 5 Page 1 of 10
<< 10 CFR 50.54(q) Effectiveness Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: Revised EREG 02310566 02303141 CNS CR3 This is a revised EREG to complete actions identified in an AFI from 2020 CNS NRC EP inspection readiness assessment. Original screen did not HNP require an evaluation.
MNS 5AD #: 02297414 ONS RNP GO Document and Revision CSD-EP-CNS-0101-01, EAL Technical Basis Document, Rev 0 I
Part I. Description of Proposed Change:
Overall changes that did not screen out and will be evaluated:
Added additional clarification to EAL Technical basis document as specified in approved NRC EPFAQs and Certificate of Compliance for Spent Fuel Storage Casks (see table below).
FPB RCS Loss Threshold Basis - EPFAQ 2018-01 CA6.1 and SA9.1 basis - EPFAQ 2018-04 Update EU1.1 to include a table that addresses both models of Spent Fuel Storage Casks in use at Catawba Nuclear Station (CNS)
Page 108: "Basis"
- Added the following paragraphs:
"An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under CA6, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under CA6, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent."
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Rev. 6 ATTACHMENT 5 Page 2 of 10
<< 10 CFR 50.54(q) Effectiveness Evaluation Form >>
Page 195: Basis
- Added the following paragraphs:
"An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under SA9, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA9 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under SA9, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent."
Page 198: "EAL"
- In "Notification of Unusual Event" box, deleted "Vertical Storage Cask (VSC) > any of the following:" and "bullets" and replaced with "spent fuel cask> any Table E-1 dose limit."
- New E-1 dose limit table added.
Page 223: "Basis" Added "If EOPs direct operators to open the Pressurizer pressure relief valves to implement a core cooling strategy (i.e., a feed and bleed cooldown), then there will exist a reactor coolant flow path from the RCS, past the pressurizer safety and relief valves and into the containment that operators cannot isolate without compromising the effectiveness of the strategy (i.e., for the strategy to be effective, the valves must be kept in the open position); therefore, the flow through the pressure relief line is UNISOLABLE. In this case, the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been eliminated and thus this condition constitutes a loss of the RCS barrier." , 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Yes Bases Validation and Verification (V&V) Form, is attached (required for IC or EAL change) No Part II. Description and Review of Licensing Basis Affected by the Proposed Change:
Licensing Basis Catawba Emergency Plan Revision 2 (dated January 1983), Revision 97-1 (dated January 1997), and Revision 11-2 (dated September 2011)
NEI 99-01 Rev. 6 EAL SER, SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 279 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NPF-52 DUKE Printed : 01/15/2020 Page : 4
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
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<< 10 CFR 50.54(q) Effectiveness Evaluation Form >>
ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 Current Emergency Plans Catawba Nuclear Station Emergency Plan, Section D, Rev. 148 The differences in approved revisions and the current revisions of the Emergency Plans have been reviewed, and they have been determined to meet the regulatory requirements required during the course of revisions.
Part Ill. Description of How the Proposed Change Complies with Regulation and Commitments.
If the emergency plan, modified as proposed, no longer complies with planning standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50, then ensure the change is rejected, modified, or processed as an exemption request under 10 CFR 50.12, Specific Exemptions, rather than under 10 CFR 50.54(q):
The change to Page 108 and 195 of Basis - Adds 3 clarifying paragraphs to EAL CA6.1 and EAL SA9.1 Three additional EAL clarifying basis information paragraphs were added to the EAL CA6.1 and SA9.1 basis as a result of NRC EPFAQ 2018-04. Industry operating experience indicated that additional clarification is needed for the following three cases:
- 1. The event affects equipment common to two or more safety systems or safety system trains.
- 2. The event affects a safety system that has only one train.
- 3. The event affects two trains of a safety system having more than two train.
The three additional paragraphs were added to address these cases. The additions provide clarification of expected emergency classification for cases not explicitly addressed by ICs CA6 and SA9; therefore, implementation of these additional paragraphs will improve accuracy and timeliness of a classification following a hazardous event affecting a safety system. The additions would result in EAL interpretations that are consistent with the meaning and intent of NRC-approved EAL bases such that the classification of the event would not be different from that approved by the NRC.
IC CA6 and SA9 is intended to be declared when actual or potential substantial performance issues with SAFETY SYSTEMS have occurred as a result of a hazardous event. Nuclear power plant SAFETY SYSTEMS are suitably redundant to assure that the safety system function can be accomplished, assuming a single failure. As such, a single failure of a SAFETY SYSTEM due to a hazardous event should not result in the declaration of an Alert. The threshold values for CA6 and SA9 meet this intention by requiring a hazardous event that causes degraded performance on one train of a SAFETY SYSTEM concurrent with either degraded performance or sufficient visible damage to a second train of a safety system to cause concern regarding the reliability or operability of the affected component. Requiring degraded performance in one SAFETY SYSTEM concurrent with either degraded performance or visible damage of a second SAFETY SYSTEM ensures that a declaration of an Alert will only be made when actual or potential substantial performance issues with SAFETY SYSTEMS have occurred as a result of a hazardous event.
The three additional paragraphs are limited to providing clarification for certain conditions and does not propose alteration of previous guidance.
This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting Printed : 01/15/2020 Page : 5
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the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels in use. The CNS Emergency Plan will continue to comply with 10 CFR 50.47(b)(4)
Emergency Classification System and 10 CFR Part 50 Appendix E, Section IV.B. Assessment Actions. Thus, an Emergency Classification System using a standard scheme of emergency classification and action levels continues to be maintained.
The change to page 198 EAL Revises EAL EU1.1 to include a table that addresses both models of Spent Fuel Storage Casks in use at Catawba Nuclear Station (CNS)
CNS has 2 models of spent fuel storage casks on-site. Each model has its own Certificate of Compliance for Spent Fuel Storage Casks with separate Technical specifications. EAL EU1.1 threshold values are based on damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific technical specification allowable radiation level) on the surface of the spent fuel cask.
Magnastor system Technical Specification allowable radiation level per 3.3.1:
The maximum surface dose rates for the CONCRETE CASK, Reference Figure A3-1, shall not exceed the following limits:
NAC-UMS system Technical Specification allowable radiation level per 3.2.2:
The average surface dose rates of each CONCRETE CASK shall not exceed the following limits unless required ACTIONS A.1 and A.2 are met
- a. 50mrem/hour (neutron + gamma) on the side (on the concrete surfaces);
- b. 50 mrem/hour (neutron + gamma) on the top;
- c. 100 mrem/hour (neutron +gamma) at air inlets and outlets New Table E-1 used in determining EAL thresholds for EU1.1 which are 2 times the Technical Specification limits.
- 240 mrem/hr (gamma) 011 the vertical
- 100 mremlh r {neutron+ gamma) on concrete surfaces the side (on the concrete surfaces}
- 1o mrem/hr (neutron) on the verti,cal . , 100 mremlh r (neutron+ gamma) on concrete surfaces th e top
- 900 mre m/hr (neutron + gamma) on the
- 200 mremlh r (neutron+ gamma) at aiir top inlets andi outlets Printed : 01/15/2020 Page : 6
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This change assures that the EAL limit is maintained in accordance with the defined basis of the EAL. This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels in use. The CNS Emergency Plan will continue to comply with 10 CFR 50.47(b)(4) Emergency Classification System and 10 CFR Part 50 Appendix E, Section IV.B. Assessment Actions. Thus, an Emergency Classification System using a standard scheme of emergency classification and action levels continues to be maintained.
The change to Page 223: "Basis" - Adds a clarifying paragraph to EAL An additional EAL clarifying basis information paragraph was added to the EAL Basis as a result of NRC EPFAQ 2018-01. This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS).
Industry operating experience indicated that additional clarification is needed for the following condition:
During a significant and protracted loss of feedwater to the steam generators, PWR emergency operating procedures (EOPs) may direct the opening of Pressurizer pressure relief valves as one action to implement a core cooling strategy often referred as a feed and bleed cooldown. The action will allow reactor coolant to exit the RCS through a pressurizer pressure relief line, collect in a pressurizer relief tank until the rupture of an engineered pressure relief device (such as a disk), and then be recirculated via the containment sumps. If there is a significant and protracted loss of feedwater to the steam generators, and EOPs direct operators to open the Pressurizer pressure relief valves to implement a core cooling strategy (i.e., a feed and bleed cooldown), then should this condition be considered a LOSS of the RCS Barrier?
The additional paragraph was added to address the above condition. The addition provides clarification of expected emergency classification for a condition not explicitly addressed by RCS loss threshold basis; therefore, implementation of this additional paragraph will improve accuracy and timeliness of a classification following the described condition. The addition would result in EAL interpretations that are consistent with the meaning and intent of NRC-approved EAL bases such that the classification of the event would not be different from that approved by the NRC.
If EOPs direct operators to open the Pressurizer pressure relief valves to implement a core cooling strategy (i.e., a feed and bleed cooldown), then there will exist a reactor coolant flow path from the RCS, past the pressurizer safety and relief valves and into the containment that operators cannot isolate without compromising the effectiveness of the strategy (i.e., for the strategy to be effective, the valves must be kept in the open position);
therefore, the flow through the pressure relief line is UNISOLABLE. In this case, the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been eliminated and thus this condition constitutes a loss of the RCS barrier.
This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels in use. The CNS Emergency Plan will continue to comply with 10 CFR 50.47(b)(4)
Emergency Classification System and 10 CFR Part 50 Appendix E, Section IV.B. Assessment Actions. Thus, an Emergency Classification System using a standard scheme of emergency classification and action levels continues to be maintained.
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The differences in approved revisions and the current revisions of the Emergency Plans have been reviewed, and continue to comply with regulations and commitments described in 10 CFR 50.47(b) and NRC requirements, as described in 10 CFR 50, Appendix E.
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<< 10 CFR 50.54(q) Effectiveness Evaluation Form >>
Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change (Address each function identified in Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV of associated Screen):
10 CFR 50.47(b)(4) states: 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 emergency planning function associated with 10 CFR 50.47(b)(4) states:
A standard scheme of emergency classification and action levels is in use.
Supporting requirements which are described in 10 CFR 50, Appendix E state:
IV.B:
- 1. The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring. By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC.
Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis.
IV.C:
- 1. The entire spectrum of emergency conditions that involve the alerting or activating of progressively larger segments of the total emergency organization shall be described. The communication steps to be taken to alert or activate emergency personnel under each class of emergency shall be described. Emergency action levels (based not only on onsite and offsite radiation monitoring information but also on readings from a number of sensors that indicate a potential emergency, such as the pressure in containment and the response of the Emergency Core Cooling System) for notification of offsite agencies shall be described. The existence, but not the details, of a message authentication scheme shall be noted for such agencies. The emergency classes defined shall include:
(1) Notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency. These classes are further discussed in NUREG-0654/FEMA-REP-1.
- 2. By June 20, 2012, nuclear power reactor licensees shall 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 shall promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. Licensees shall not construe these criteria as a grace period to attempt to restore plant conditions to avoid declaring an emergency action due to an emergency action level that has been exceeded. Licensees shall not construe these criteria as Printed : 01/15/2020 Page : 9
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preventing implementation of response actions deemed by the licensee to be necessary to protect public health and safety provided that any delay in declaration does not deny the State and local authorities the opportunity to implement measures necessary to protect the public health and safety.
The applicable program elements described in NUREG-0654,Section II.D state:
D.1: An emergency classification and emergency action level scheme as set forth in Appendix 1 must be established by the licensee. The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures. The plan shall identify the parameter values and equipment status for each emergency class.
D.2: The initiating conditions shall include the example conditions found in Appendix 1 and all postulated accidents in the Final Safety Analysis Report (FSAR) for the nuclear facility.
Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:
The change to Page 108 and 195 of Basis - Adds 3 clarifying paragraphs to EAL CA6.1 and EAL SA9.1 Three additional EAL clarifying basis information paragraphs were added to EALs CA6.1 and SA9.1 basis as a result of NRC EPFAQ 2018-04. This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels is in use. Thus, the Effectiveness of Emergency Plan Function for the Emergency Classification System is not reduced.
The change to page 198 EAL Revises EAL EU1.1 to include a table that addresses both models of Spent Fuel Storage Casks in use at Catawba Nuclear Station (CNS)
The NAC Magnastor EAL thresholds are based on two times the applicable Technical Specification limits for dose rates on the NAC Magnastor loaded canisters. This change assures that the EAL limit is maintained in accordance with the defined basis of the EAL. This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels in use. Thus, the Effectiveness of Emergency Plan Function for the Emergency Classification System is not reduced.
The change to Page 223: "Basis" - Adds a clarifying paragraph to EAL An additional EAL clarifying basis information paragraph were added to the EAL Basis as a result of NRC EPFAQ 2018-01. This change is an enhancement to the basis to assure proper and consistent application of the EAL without affecting the timelines or accuracy of the classification. This clarification does not affect the standard scheme of emergency classification and action levels in use. Thus, the Effectiveness of Emergency Plan Function for the Emergency Classification System is not reduced.
The proposed changes in Revision 0 of CSD-EP-CNS-0101-01 do not reduce the effectiveness of the Catawba Nuclear Station Emergency Plan. Instead, these changes continue to provide additional assurance that the Printed : 01/15/2020 Page : 10
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Emergency Response Organization has the ability and capability to:
- respond to an emergency;
- perform functions in a timely manner;
- effectively identify and take measures to ensure protection of the public health and safety; and
- effectively use response equipment and emergency response procedures.
The proposed changes in Revision 0 of CSD-EP-CNS-0101-01 continue to support a classified emergency, resulting in an improved capability to ensure health and safety of plant personnel and the general public. These changes continue to meet NRC requirements, as described in 10 CFR 50.47(b) and 10 CFR 50, Appendix E as well as the requirements of the Duke Energy Sites Emergency Plans as written and approved.
Part VI. Evaluation Conclusion.
Answer the following questions about the proposed change.
1 Does the proposed change comply with 10 CFR 50.47(b) and 10 CFR 50 Appendix E? Yes No 2 Does the proposed change maintain the effectiveness of the emergency plan (i.e., no Yes No reduction in effectiveness)?
3 Does the proposed change maintain the current Emergency Action Level (EAL) scheme? Yes No 4 Choose one of the following conclusions:
a The activity does continue to comply with the requirements of 10 CFR 50.47(b) and 10 CFR 50, Appendix E, and the activity does not constitute a reduction in effectiveness or change in the current Emergency Action Level (EAL) scheme. Therefore, the activity can be implemented without prior NRC approval.
b The activity does not continue to comply with the requirements of 10 CFR 50.47(b) or 10 CFR 50 Appendix E or the activity does constitute a reduction in effectiveness or EAL scheme change.
Therefore, the activity cannot be implemented without prior NRC approval.
Part VII. Disposition of Proposed Change Requiring Prior NRC Approval Will the proposed change determined to require prior NRC approval be either revised or Yes No rejected? I I If No, then initiate a License Amendment Request in accordance 10 CFR 50.90, AD-LS-ALL-0002, Regulatory Correspondence, and AD-LS-ALL-0015, License Amendment Request and Changes to SLC, TRM, and TS Bases, and include the tracking number: .
Part VIII. Signatures: EP CFAM Final Approval is required for changes affecting risk significant planning standard 10 CFR 50.47(b)(4) (i.e., Emergency Action Levels and Emergency Action Level Bases). If CFAM approval is NOT required, then mark the CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Matthew Nelson CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
See CAS CAS See CAS Printed : 01/15/2020 Page : 11
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Approver (Manager, Nuclear Support Approver Signature: Date:
Services) Name (Print): CAS See CAS See CAS Approver (CFAM, as required) Name (Print): Approver Signature: Date:
See CAS See CAS See CAS If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General Assignments.
If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 01/15/2020 Page : 12
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02303180 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02303180 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 12/12/2019 Pri Resp Fac :
Status : COMPLETE Reschedule :0 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 18255 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes approve changes Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 11/18/2019 I44004 INPROG 11/18/2019 I44004 12/12/2019 11/18/2019 I44004 NTFY/ASG 11/18/2019 I44004 ACC/ASG 11/18/2019 I44004 AWAIT/C 11/21/2019 MRCOYL1 ACC/ASG 11/21/2019 I44004 AWAIT/C 12/16/2019 MEHARE COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By check boxes for 30 day submittal 20191121 MRCOYL1 Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02303180 Duke Energy ASSIGNMENT NBR - 01 Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MRCOYL1 A 11/21/2019 1136 APPROVED 11/21/2019 1207 COYLE MRCOYL1 A 11/18/2019 1624 RETURNED 11/21/2019 0923 COYLE MEHARE A 11/21/2019 1207 APPROVED 12/16/2019 1800 HARE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :2
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<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02303180 CNS CR3 HNP MNS 5AD #: 02298688 ONS RNP GO Document and Revision I EPA E, Notification Methodology, Rev 150 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
Page E-1:
- E.2 - Removed "Catawba Emergency Response Procedures RP/0/A/5000/006A, Notification to States and Counties from the Control Room, and"
- E.2.a Note - Removed "and Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/001, Classification of Emergency" Page E-2:
- E.2.b Note - Removed "and Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/001, Classification of Emergency" Page E-3:
- 6th paragraph - Removed "RP/0/A/5000/003, Alert, or" Page E-4:
- E.2.c Note - Removed "and Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/001, Classification of Emergency" Page E-5:
- 2nd paragraph - Removed "RP/0/A/5000/004, Site Area Emergency, or"
- 6th paragraph - Removed entire paragraph "The Emergency Coordinator will assure notification of all Catawba Nuclear Station management not notified thus far for those initiating conditions or implementation of any Emergency Procedure affecting these personnel in accordance with Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/004, Site Area Emergency or AD-EP-ALL-0101, Emergency Classification" Page E-6:
- E.2.d Note - Removed "and Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/001, Classification of Emergency"
- E.2.d, 6th paragraph - Removed "RP/0/A/5000/005, General Emergency or" Page E-7:
- Step 5, 1st paragraph - Removed entire paragraph "The Emergency Coordinator will assure notification of all Catawba Nuclear Station management not notified thus far for those initiating conditions or implementation of any Emergency Procedure affecting these personnel in accordance with Catawba Nuclear Station Emergency Response Procedure RP/0/A/5000/005, General Emergency or AD-EP-ALL-0101, Emergency Printed : 12/17/2019 Page : 3 Back
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- Step 5, 2nd paragraph - Removed "RP/0/A/5000/005, General Emergency, or" Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No I 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional) I Part III. Editorial Change No or Is this activity an editorial or typographical change only, such as formatting, Yes I
Partially I paragraph numbering, spelling, or punctuation that does not change intent? 10 CFR 50.54(q) Continue to Effectiveness Attachment 4, Evaluation is not Part IV and Justification:
required. Enter address non justification and editorial The changes to the procedure titles are editorial in nature in accordance with AD-complete changes EP-ALL-0602, step 3.0.6.g, "Correct references that are no longer applicable."
Attachment 4, Fleet Document AD-EP-ALL-0111, Control Room Activation of the ERO, is Part V.
superseding the following procedures:
o RP/0/A/5000/001, Classification of Emergency o RP/0/A/5000/002, Notification of Unusual Event o RP/0/A/5000/003, Alert o RP/0/A/5000/004, Site Area Emergency o RP/0/A/5000/005, General Emergency o RP/0/A/5000/006 A, Notifications to States and Counties from the Control Room Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
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2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that on shift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of on shift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures 5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3)
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Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b) (10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b) (11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b) (12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b) (13) Recovery Planning and Post-Accident Operations 13a Plans for recovery and reentry are developed.
14 10 CFR 50.47(b) (14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 5 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part IV. Emergency Planning Element and Function Screen (cont.)
15 10 CFR 50.47(b) (15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b) (16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
The non-editorial change that removed the paragraph in which the Emergency Coordinator made notification to non-ERO Catawba Nuclear Station Management does not affect a planning standard. This was a courtesy notification and not related to the Emergency Plan.
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete , 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle See CAS See CAS Approver (Manager, Nuclear Support Services) Approver Signature: Date:
Name (Print): See CAS See CAS Mandy Hare Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A Printed : 12/17/2019 Page : 7
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 6 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 12/17/2019 Page : 8
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02267546 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02267546 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 12/31/2019 Pri Resp Fac :
Status : COMPLETE Reschedule :1 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 13650 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 04/09/2019 I44004 INPROG 04/09/2019 I44004 04/24/2019 04/09/2019 I44004 NTFY/ASG 04/09/2019 I44004 ACC/ASG 06/10/2019 I44004 12/31/2019 11/18/2019 I44004 AWAIT/C 11/21/2019 MRCOYL1 ACC/ASG 11/21/2019 I44004 AWAIT/C 12/16/2019 MEHARE ACC/ASG 12/16/2019 I44004 AWAIT/C 12/16/2019 MRCOYL1 ACC/ASG 12/16/2019 I44004 AWAIT/C 12/16/2019 MEHARE COMPLETE Action Request Assignment Routing/Return Comments Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02267546 Duke Energy ASSIGNMENT NBR - 01 Routing Comments from the X601 Panel Updated On Updated By 30 day submittal section isn't included on the document but should be 20191121 MRCOYL1 checked. 20191121 MRCOYL1 Need item updated in I.1 on Q to reflect current wording in section. 20191216 MEHARE
. 20191216 MEHARE Rev number 20191216 MRCOYL1 Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MRCOYL1 A 12/16/2019 1826 APPROVED 12/16/2019 1831 COYLE MRCOYL1 A 11/18/2019 1637 RETURNED 11/21/2019 1129 COYLE MEHARE A 11/21/2019 1319 RETURNED 12/16/2019 1707 HARE MRCOYL1 A 11/21/2019 1153 APPROVED 11/21/2019 1319 COYLE MEHARE A 12/16/2019 1831 APPROVED 12/16/2019 1847 HARE MRCOYL1 A 12/16/2019 1730 RETURNED 12/16/2019 1820 COYLE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Printed : 12/17/2019 Page :2
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02267546 ASSIGNMENT NBR - 01 Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :3
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 1 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02267546 CNS CR3 HNP MNS ONS 5AD #: 02267543 RNP GO Document and Revision Emergency Plan Section I, Accident Assessment, Rev 148 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
E-Plan or Current (Existing) Text Proposed (Change) Text Procedure Section Reference I.1 RP/0/A/5000/001, Classification of Emergency None and AD-EP-ALL-0101, Emergency AD-EP-ALL-0101, Emergency Classification I.1 NEI 99-01 Revision 6 Wallboard CSD-EP-CNS-0101-02, EAL Wall Charts I.3.A/I.3.B AD-EP-CNS-0202 AD-EP-CNS-0203 I.7/I.8 AD-EP-CNS-0203, Site Specific Field AD-EP-CNS-0203, CNS Site Specific Monitoring Information Field Monitoring I.10 Provisions for assessing contamination levels, Provisions for assessing contamination water, and air to dose rates for key isotopes is levels, water, and air to dose rates for found in procedure HP/0/B/1009/024. key isotopes is found in the Offsite Dose Calculation Manual (ODCM)
Printed : 12/17/2019 Page : 4 Back
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 2 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No I 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional) I Part III. Editorial Change No or Yes I
Partially I Printed : 12/17/2019 Page : 5
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 3 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Is this activity an editorial or typographical change only, such as formatting, 10 CFR 50.54(q) Continue to paragraph numbering, spelling, or punctuation that does not change intent? Effectiveness Attachment 4, Evaluation is not Part IV and required. Enter address non Justification: justification and editorial All changes in this revision are editorial, based on the guidance in AD-EP-ALL- complete changes 0602, Section 3.0.6, Editorial Change. Attachment 4, Part V.
The change to delete, "RP/0/A/5000/001, Classification of Emergency" is a change to delete a superseded procedure. (step 3.0.6.g.)
The changes in section I.1 from "AD-EP-ALL-0101, Emergency" to,"AD-EP-ALL-0101, Emergency Classification" and "NEI 99-01 Revision 6 EAL Wallboard" to "CSD-EP-CNS-0101-02 Wallcharts" are changes to correct reference to steps, pages, attachments, forms, documents, tables, exhibits and procedures. (step 3.0.6.b)
The change in section I.3.A/I.3.B from "AD-EP-CNS-0202" to "AD-EP-CNS-0203" is a change to correct reference to steps, pages, attachments, forms, documents, tables, exhibits and procedures. (step 3.0.6.b). AD-EP-CNS-0202 does not exist.
The change in section I.7/I.8 from "AD-EP-CNS-0203, Site Specific Field Monitoring Information" to "AD-EP-CNS-0203, CNS Site Specific Field Monitoring" is a change to correct reference to steps, pages, attachments, forms, documents, tables, exhibits and procedures. (step 3.0.6.b).
The change in section I.10 from, "Provisions for assessing contamination levels, water, and air to dose rates for key isotopes is found in procedure HP/0/B/1009/024" to " Provisions for assessing contamination levels, water, and air to dose rates for key isotopes is found in the Offsite Dose Calculation Manual (ODCM)" is a change to correct references or annotations that are no longer applicable (step 3.0.6.g). PRR 02230283 deleted HP/0/B/1009/024 and EREG 022300502 evaluated the deletion of HP/0/B/1009/024.
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2aI Process ensures that onshift emergency response responsibilities are staffed and assigned Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 4 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
2b The process for timely augmentation of onshift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan.
(NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use.
(Requires final approval of Screen and Evaluation by EP CFAM.)
Part IV. Emergency Planning Element and Function Screen (cont.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures 5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3) 6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b)(10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11IPrinted 10 CFR 50.47(b)(11) Radiological Exposure Control
- 12/17/2019 Page : 7
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 5 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b)(12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b)(13) Recovery Planning and Post-accident Operations 13a Plans for recovery and reentry are developed.
Part IV. Emergency Planning Element and Function Screen (cont.)
14 10 CFR 50.47(b)(14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
15 10 CFR 50.47(b)(15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b)(16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete , 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle See CAS See CAS Approver (EP Manager Name (Print): Approver Signature: Date:
Mandy Hare See CAS See CAS Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A Printed : 12/17/2019 Page : 8
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 6 of 6
<< 10 CFR 50.54(q) Screening Evaluation Form >>
If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 12/17/2019 Page : 9
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02299062 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02299062 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 01/30/2020 Pri Resp Fac :
Status : COMPLETE Reschedule :1 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 18255 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes I reviewed the 50.54Q and PRRA for EPA J Rev. 148 and approve the changes.
Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 10/24/2019 I44004 INPROG 10/24/2019 I44004 11/14/2019 10/24/2019 I44004 NTFY/ASG 10/24/2019 I44004 ACC/ASG 12/10/2019 I44004 01/30/2020 12/11/2019 I44004 AWAIT/C 12/15/2019 MEHARE COMPLETE Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By
- No Routing Comments Found ***
Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02299062 Duke Energy ASSIGNMENT NBR - 01 Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MRCOYL1 A 12/11/2019 0906 APPROVED 12/12/2019 1008 COYLE MEHARE A 12/12/2019 1008 APPROVED 12/15/2019 2039 HARE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :2
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 1 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02299062 CNS CR3 HNP MNS 5AD #: 02245206 ONS RNP GO Document and Revision EMERGENCY PLAN SECTION J, PROTECTIVE RESPONSE, REV 148 I
(DRR 02161925)
Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
Page J-3: J.7 Protective Action Recommendations
- Removed RP/0/A/5000/005, General Emergency Page J-6: J.10.a EPZ Maps
- Changed Figure J-4 to J-3.
Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional)
Printed : 12/17/2019 Page : 3 Back
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 2 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
I I Part III. Editorial Change No or Yes I I Is this activity an editorial or typographical change only, such as formatting, Partially paragraph numbering, spelling, or punctuation that does not change intent? 10 CFR 50.54(q) Continue to Effectiveness Attachment 4, Evaluation is not Part IV and Justification:
required. Enter address non justification and editorial The change to remove RP/0/A/5000/005, General Emergency, is editorial complete changes based on AD-EP-ALL-0602. step 3.0.g, "Correct references or annotations that Attachment 4, are no longer applicable." RP/0/A/5000/005 has been superseded.
Part V.
The change to correct the figure title from Figure J-4 to J-3 is editorial based on AD-EP-ALL-0602. step 3.0.b, "Correct references to steps, pages, attachment, forms, documents tables, exhibits and procedures.
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that on shift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of on shift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures 5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3)
Printed : 12/17/2019 Page : 4
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 3 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
5c 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-I I approved ANS design report and supporting FEMA approval letter. (NA for CR3) I I Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b) (10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) 10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b) (11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b) (12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b) (13) Recovery Planning and Post-Accident Operations 13a Plans for recovery and reentry are developed.
14 10 CFR 50.47(b) (14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
Printed : 12/17/2019 Page : 5
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 4 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
14b 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.
I14c I Identified weaknesses are corrected. I I
Part IV. Emergency Planning Element and Function Screen (cont.)
15 10 CFR 50.47(b) (15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b) (16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Ryder Coyle See CAS See CAS Approver (Manager, Nuclear Support Services) Approver Signature: Date:
Name (Print): See CAS See CAS Mandy Hare Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AD-EP-ALL-0602 AND EFFECTIVENESS EVALUATIONS 10 Rev. 6 CFR 50.54(Q)
ATTACHMENT 4 Page 5 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form >>
If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 12/17/2019 Page : 7
- END OF REPORT***
2019-12-17 07:44:32.438
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02267856 ASSIGNMENT NBR - 01 Action Request Assignment Details AR NUMBER : 02267856 ASSIGNMENT NUMBER : 01 Type : EP01 Due Date : 12/19/2019 Pri Resp Fac :
Status : COMPLETE Reschedule :3 Pri Resp Group :
Assigned To : STACI N WHITE Sec Resp Fac :
Subject : 50.54(Q) SCREEN Sec Resp Group :
Aff Facility : CN Unit : System :
UCR : Schedule Ref :
Organization : Department : 13650 Discipline :
Est Manhrs :0 Est Comp Date :
Description COMPLETE 50.54(Q) SCREEN IN ACCORDANCE WITH AD-EP-ALL-0602.
Action Request Assignment Completion Notes approve changes in 50.54Q for EPA P Rev 150 Action Request Assignment Status History Updated Date Updated By Assgn Status Assgn Due Date 04/10/2019 I44004 INPROG 04/10/2019 I44004 04/24/2019 04/10/2019 I44004 NTFY/ASG 04/10/2019 I44004 ACC/ASG 06/10/2019 I44004 08/30/2019 09/30/2019 I44004 10/31/2019 10/22/2019 I44004 12/19/2019 12/11/2019 I44004 AWAIT/C 12/12/2019 MRCOYL1 ACC/ASG 12/12/2019 I44004 AWAIT/C 12/16/2019 MEHARE ACC/ASG 12/16/2019 I44004 AWAIT/C 12/16/2019 MRCOYL1 ACC/ASG 12/16/2019 I44004 AWAIT/C 12/16/2019 MEHARE COMPLETE Printed : 12/17/2019 Page :1
I (~ DUKE ENERGY, ACTION REQUEST - 02267856 Duke Energy ASSIGNMENT NBR - 01 Action Request Assignment Routing/Return Comments Routing Comments from the X601 Panel Updated On Updated By Address the EPIPs that are being added to this section. 20191212 MRCOYL1 Update 50.54Q to address all items in revision summary. 20191216 MEHARE E-plan boxes need to be checked. 20191216 MRCOYL1 Routing Comments from the X602 Panel Updated On Updated By
- No Return Comments Found ***
Action Request Assignment Completion Approval Route List : 001 Route List Initiator : I44004 Send Send Action Action Passport Fac Group / Type Date Time Taken Date / Time Last Name MEHARE A 12/16/2019 1834 APPROVED 12/16/2019 1858 HARE MRCOYL1 A 12/16/2019 1814 RETURNED 12/16/2019 1827 COYLE MRCOYL1 A 12/16/2019 1828 APPROVED 12/16/2019 1834 COYLE MRCOYL1 A 12/11/2019 0913 RETURNED 12/12/2019 1511 COYLE MEHARE A 12/12/2019 1539 RETURNED 12/16/2019 1710 HARE MRCOYL1 A 12/12/2019 1525 APPROVED 12/12/2019 1539 COYLE Action Request Assignment Cause/Action Action Request Assignment Reference Documents Doc Sub Minor Facility Type Type Document Sheet Rev Rev Title Action Request Assignment Reference Equipment Equip Equip Equip Equip Rev Facility Unit System Type Number Tag Status Rev Status Printed : 12/17/2019 Page :2
I (~ DUKE ENERGY, Duke Energy ACTION REQUEST - 02267856 ASSIGNMENT NBR - 01 Action Request Assignment Cross References Ref Ref Ref Ref Nbr Limit Type Nbr Sub Type Status AS Cls Description Action Request Assignment Appendices APPENDIX 1 Printed : 12/17/2019 Page :3
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 1 of 8
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Screening and Evaluation Number Applicable Sites BNP EREG #: 02267856 CNS CR3 HNP MNS 5AD #: 02267855 ONS RNP GO Document and Revision Emergency Plan Section P, Responsibility for the Planning Effort, Rev 150 Printed : 12/17/2019 Page : 4 Back
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
E-Plan or Current (Existing) Text Proposed (Change) Text Procedure Section Reference P.9 The Nuclear Safety Review Board The Nuclear Oversight Manager Nuclear QA Audits will arrange for an independent review...
Chairman will arrange for an independent review...
P.9 none Guidance for performing the assessment against the performance indicators is provided in Emergency Preparedness Administrative Procedure AD-EP-ALL-0001.
P.9 The independent review will be The independent review will be conducted by the Nuclear Oversight Division which will include conducted by the Nuclear Oversight-Audits and will include Figure P-2, Procedure # Title Emergency Plan Section Implemented Emergency Plan Implementing Procedures none AD-EP-ALL-0100 Emergency Response Organization (ERO) Section A, B AD-EP-ALL-0111 Control Room Activation of the ERO Section D, E, J, K AD-EP-ALL-0500 Emergency Response Training Section O AD-EP-ALL-0501 Emergency Preparedness Staff Training and Qualification Section P AD-EP-ALL-0801 Design and Development of Drill and Exercises Section N AD-EP-ALL-0802 Conducting Drills and Exercises Section N AD-EP-ALL-0803 Evaluation and Critique of Drills and Exercises Section N HP/0/B/1000/010 Determination of Radiation Monitor Setpoints Section D AD-EP-ALL-0109, Protective Action AD-EP-ALL-0109, Offsite Protective Action Recommendations Recommendations HP/0/B/1009/008, Contamination HP/0/B/1009/008, Contamination Control of Injured Individuals, Section K.5, L.1, L.4 Control of Injured Individuals, Section (REMOVED SECTION D)
D, K.5, L.1, L.4 Printed : 12/17/2019 Page : 5
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 3 of 8
<< 10 CFR 50.54(q) Screening Evaluation Form >>
HP/0/B/1009/003, Radiation Deleted Protection Response Following a Primary to Secondary Leak HP/0/B/1009/024, Implementing Deleted Procedure for Estimating Food Chain Doses Under Post-Accident Conditions,Section I.9 RP/0/A/5000/001, Classification of Deleted Emergency RP/0/A/5000/002, Notification of Deleted Unusual Event RP/0/A/5000/003, Alert Deleted RP/0/A/5000/004, Site Area Deleted Emergency RP/0/A/5000/005, General Deleted Emergency RP/0/A/5000/06 A, Notifications to Deleted States and Counties from the Control Room Printed : 12/17/2019 Page : 6
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 Part II. Activity Previously Reviewed?
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Yes I No I 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4, Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part III Justification: Attachment 4, Part V.
Bounding document attached (optional) I Part III. Editorial Change No or Is this activity an editorial or typographical change only, such as formatting, Yes I
Partially I paragraph numbering, spelling, or punctuation that does not change intent? 10 CFR 50.54(q) Continue to Effectiveness Attachment 4, Evaluation is not Part IV and Justification:
required. Enter address non The change from Nuclear Safety Review Board (NSRB) Chairman to Nuclear justification and editorial Oversight Manager Nuclear QA Audits is editorial because it is a change to complete changes correct the title with no change in the authority or responsibility. (AD-EP-ALL- Attachment 4, 0602, Section 3.0.6.d). The role of the NSRB has changed. The Chairman and Part V.
the NRSB no longer have the responsibility for the Audit Program, that is now the Nuclear Oversight Manager.
The change from Nuclear Oversight-Audits to Nuclear Oversight Division is editorial because it is a change to correct the title with no change in the authority or responsibility. (AD-EP-ALL-0602, Section 3.0.6.d).
The change to add AD-EP-ALL-0100, AD-EP-ALL-0109, AD-EP-ALL-0111, AD-EP-ALL-0500, AD-EP-ALL-0501, AD-EP-ALL-0801, AD-EP-ALL-0802, AD-EP-ALL-0803, and HP/0/B/1000/010 is editorial because it corrects references to procedures that implement the Emergency Plan. (AD-EP-ALL-0602, section 3.0.6.b)
The change to delete HP/0/B/1009/003, HP/0/B/1009/024, RP/0/A/5000/001, RP/0/A/5000/002, RP/0/A/5000/003, RP/0/A/5000/004, RP/0/A/5000/005, and RP/0/A/5000/06 A is editorial because it corrects references to procedures that have been superseded. (AD-EP-ALL-0602, section 3.0.6.g)
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
IPrinted : 12/17/2019 Page : 7
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 5 of 4
<< 10 CFR 50.54(q) Screening Evaluation Form >>
1 10 CFR 50.47(b)(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
1b 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.
2 10 CFR 50.47(b)(2) Onsite Emergency Organization 2a Process ensures that onshift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of onshift staff is established and maintained.
3 10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
3b State and local staff can be accommodated at the EOF in accordance with the emergency plan.
(NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use.
(Requires final approval of Screen and Evaluation by EP CFAM.)
Part IV. Emergency Planning Element and Function Screen (cont.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures 5a Procedures for notification of State and local governmental agencies are capable of alerting them of the declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
5b Administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway. (NA for CR3) 5c 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. (NA for CR3) 6 10 CFR 50.47(b)(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response organizations.
6b Systems are established for prompt communication to emergency response personnel.
7 10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
8 10 CFR 50.47(b)(8) Emergency Facilities and Equipment 8a Adequate facilities are maintained to support emergency response.
8b Adequate equipment is maintained to support emergency response.
9 10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
10 10 CFR 50.47(b)(10) Protective Response 10a A range of public PARs is available for implementation during emergencies. (NA for CR3)
Printed : 12/17/2019 Page : 8
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 6 of 4
<< 10 CFR 50.54(q) Screening Evaluation Form >>
10b 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. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including those for hostile action events.
10d KI is available for implementation as a protective action recommendation in those jurisdictions that chose to provide KI to the public.
11 10 CFR 50.47(b)(11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
12 10 CFR 50.47(b)(12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
13 10 CFR 50.47(b)(13) Recovery Planning and Post-accident Operations 13a Plans for recovery and reentry are developed.
Part IV. Emergency Planning Element and Function Screen (cont.)
14 10 CFR 50.47(b)(14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) is established.
14b 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.
14c Identified weaknesses are corrected.
15 10 CFR 50.47(b)(15) Emergency Response Training 15a Training is provided to emergency responders.
16 10 CFR 50.47(b)(16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
16b Planners responsible for emergency plan development and maintenance are properly trained.
PART IV. Conclusion If no Part IV criteria are checked, then provide Justification and complete Part V below.
Justification:
The following changes are not editorial, as defined in AD-EP-ALL-0602, nor do they affect a Planning Printed : 12/17/2019 Page : 9
EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
Rev. 6 ATTACHMENT 4 Page 7 of 4
<< 10 CFR 50.54(q) Screening Evaluation Form >>
Standard.
The additional wording of "Guidance for performing the assessment against the performance indicators is provided in Emergency Preparedness Administrative Procedure AD-EP-ALL-0001, Emergency Preparedness Performance Indicators" does not affect a planning standard because it is a reference to the existing procedure that governs Performance Indicators.
Removing Section D from the list of Emergency Plan Sections implemented by HP/0/B/1009/008, Contamination Control of Injured Individuals, should have been done when the NEI 99-01, rev 6 EALs were implemented 3/9/2017. The new EALs do not include Contaminated Injured Individuals in the EAL scheme.
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Preparer Name (Print): Preparer Signature: Date:
Staci White See CAS See CAS Reviewer Name (Print): Reviewer Signature: Date:
Michael Coyle See CAS See CAS Approver (EP Manager Name (Print): Approver Signature: Date:
Mandy Hare See CAS See CAS Approver (EP CFAM, as required) Name (Print): Approver Signature: Date:
N/A N/A N/A If the proposed activity is a change to the E-Plan, then initiate PRRs.
If the proposed activity is a change to the E-Plan, then create two EREG General assignments If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD Printed : 12/17/2019 Page : 10
- END OF REPORT***
2019-12-17 07:44:32.438
Catawba Nuclear Station Attachment 2 Emergency Plan Revision 20-01 Plan Update Instructions Replace Revision 19-04 Coversheet with Revision 20-01 Coversheet List of Effective Pages (LOEP)
Replace all pages of this section Tab D - Has been SUPERSEDED by CSD-EP-CNS-0101-01, EAL Technical Basis Document Replace all pages of this section Tab E - Notification Methodology Replace all pages of this section Tab I - Accident Assessment Replace all pages of this section Tab J - Protective Response Replace all pages of this section Tab P - Responsibility for the Planning Effort Replace all pages of this section
Catawba Nuclear Station Attachment 3 Emergency Plan Emergency Plan Revision 20-01
LIST OF EFFECTIVE PAGES (LOEP)
This page intentionally left blank.
Rev. 159 Page 1 of 5 January 2020
LIST OF EFFECTIVE PAGES (LOEP)
EMERGENCY PLAN SECTION PAGE NUMBER REVISION NUMBER DATE Emergency Plan Approval Cover Sheet 16-1 March 2016 Table of Contents 1 150 February 2019 2 150 February 2019 3 150 February 2019 4 150 February 2019 5 150 February 2019 6 150 February 2019 List of Figures, Tables and Attachments 7 16-1 March 2016 8 16-1 March 2016 9 16-1 March 2016 Introduction i-1 16-1 March 2016 i-2 16-1 March 2016 i-3 16-1 March 2016 i-4 16-1 March 2016 i-5 16-1 March 2016 i-6 16-1 March 2016 Section A A-1 150 September 2017 A-2 150 September 2017 A-3 150 September 2017 A-4 150 September 2017 A-5 150 September 2017 Section B B-1 167 September 2019 B-2 167 September 2019 B-3 167 September 2019 B-4 167 September 2019 B-5 167 September 2019 B-6 167 September 2019 B-7 167 September 2019 B-8 167 September 2019 B-9 167 September 2019 B-10 167 September 2019 B-11 167 September 2019 B-12 167 September 2019 B-13 167 September 2019 B-14 167 September 2019 B-15 167 September 2019 Section C C-1 16-1 March 2016 C-2 16-1 March 2016 Section D 1-258 (all pages) SUP December 2019 Section E E-1 150 December 2019 E-2 150 December 2019 Rev. 159 Page 2 of 5 January 2020
LIST OF EFFECTIVE PAGES (LOEP)
EMERGENCY PLAN SECTION PAGE NUMBER REVISION NUMBER DATE E-3 150 December 2019 E-4 150 December 2019 E-5 150 December 2019 E-6 150 December 2019 E-7 150 December 2019 E-8 150 December 2019 E-9 150 December 2019 E-10 150 December 2019 Section F F-1 16-1 March 2016 F-2 16-1 March 2016 F-3 16-1 March 2016 F-4 16-1 March 2016 F-5 16-1 March 2016 F-6 16-1 March 2016 F-7 16-1 March 2016 Section G G-1 161 February 2019 G-2 161 February 2019 G-3 161 February 2019 Section H H-1 150 October 2019 H-2 150 October 2019 H-3 150 October 2019 H-4 150 October 2019 H-5 150 October 2019 H-6 150 October 2019 H-7 150 October 2019 H-8 150 October 2019 H-9 150 October 2019 H-10 150 October 2019 H-11 150 October 2019 H-12 150 October 2019 H-13 150 October 2019 H-14 150 October 2019 Section I I-1 148 December 2019 I-2 148 December 2019 I-3 148 December 2019 I-4 148 December 2019 Section J J-1 148 December 2019 J-2 148 December 2019 J-3 148 December 2019 J-4 148 December 2019 J-5 148 December 2019 J-6 148 December 2019 J-7 148 December 2019 Rev. 159 Page 3 of 5 January 2020
LIST OF EFFECTIVE PAGES (LOEP)
EMERGENCY PLAN SECTION PAGE NUMBER REVISION NUMBER DATE J-8 148 December 2019 J-9 148 December 2019 Section K K-1 144 September 2017 K-2 144 September 2017 K-3 144 September 2017 K-4 144 September 2017 Section L L-1 11-1 May 2011 L-2 11-1 May 2011 Section M M-1 162 October 2019 M-2 162 October 2019 M-3 162 October 2019 M-4 162 October 2019 M-5 162 October 2019 Section N N-1 149 February 2019 N-2 149 February 2019 N-3 149 February 2019 Section O O-1 144 September 2017 O-2 144 September 2017 Section P P-1 150 December 2019 P-2 150 December 2019 P-3 150 December 2019 P-4 150 December 2019 P-5 150 December 2019 P-6 150 December 2019 P-7 150 December 2019 Section Q Q-1 154 November 2019 Appendix 1 Q-1.1 154 November 2019 Q-1.2 154 November 2019 Q-1.3 154 November 2019 Q-1.4 154 November 2019 Q-1.5 154 November 2019 Q-1.6 154 November 2019 Q-1.7 154 November 2019 Q-1.8 154 November 2019 Q-1.9 154 November 2019 Section Q, Appendix 2 Q-2.1 154 November 2019 Q-2.2 154 November 2019 Q-2.3 154 November 2019 Q-2.4 154 November 2019 Q-2.5 154 November 2019 Appendix 3 Q-3.1 154 November 2019 Q-3.2 154 November 2019 Rev. 159 Page 4 of 5 January 2020
LIST OF EFFECTIVE PAGES (LOEP)
EMERGENCY PLAN SECTION PAGE NUMBER REVISION NUMBER DATE Q-3.3 154 November 2019 Q-3.4 154 November 2019 Q-3.5 154 November 2019 Appendix 4 Q-4.1 154 November 2019 Appendix 5 Q-5.1 154 November 2019 Q-5.2 154 November 2019 Q-5.3 154 November 2019 Rev. 159 Page 5 of 5 January 2020
CNS Emergency Plan Section D Emergency Classification System Superseded by CSD-EP-CNS-0101-01 EAL Technical Basis Document
( ~ DUKE CATAWBA NUCLEAR STATION ENERGY EAL TECHNICAL BASIS DOCUMENT Revision 000 CSD-EP-CNS-0101-01 Rev. 000 Page 1 of 260
TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ................................................................................................................................... 3 2.0 DISCUSSION............................................................................................................................... 3 2.1 Background ............................................................................................................................... 3 2.2 Fission Product Barriers ............................................................................................................ 4 2.3 Fission Product Barrier Classification Criteria ........................................................................... 4 2.4 EAL Organization ...................................................................................................................... 5 2.5 Technical Bases Information..................................................................................................... 7 2.6 Operating Mode Applicability .................................................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................................... 9 3.1 General Considerations .......................................................................................................... 10 3.2 Classification Methodology ..................................................................................................... 10
4.0 REFERENCES
.......................................................................................................................... 14 4.1 Developmental ........................................................................................................................ 14 4.2 Implementing .......................................................................................................................... 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS .................................................................... 15 6.0 CNS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ............................................................ 24 7.0 ATTACHMENTS ........................................................................................................................ 28 1 Emergency Action Level Technical Bases .................................................................... 29 Category R Abnormal Rad Release / Rad Effluent .............................................. 29 Category C Cold Shutdown / Refueling System Malfunction................................ 71 Category H Hazards ........................................................................................... 108 Category S System Malfunction ......................................................................... 153 Category E ISFSI ............................................................................................... 197 Category F Fission Product Barrier Degradation ............................................... 200 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ......................................................................................................... 205 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases.......................... 253 CSD-EP-CNS-0101-01 Rev. 000 Page 2 of 260
1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Catawba Nuclear Station (CNS). It should be used to facilitate review of the CNS EALs and provide historical documentation for future reference.
Decision-makers responsible for Classification of Emergency may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Coordinator 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 Coordinator 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 CNS 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 3 of 260
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).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ADAMS Accession Number ML12326A805) (ref.
4.1.1), CNS 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 an increased probability of barrier loss and decreased certainty of maintaining the barrier.
The primary fission product barriers are:
A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.
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 CSD-EP-CNS-0101-01 Rev. 000 Page 4 of 260
2.4 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 NCS 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 the third barrier CSD-EP-CNS-0101-01 Rev. 000 Page 5 of 260
2.4 EAL Organization The CNS EAL scheme includes the following features:
- Division of the EAL set into three broad groups:
o EALs applicable under all 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, Hot Standby, 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 CNS EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the CNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CNS EAL categories and subcategories are listed below.
CSD-EP-CNS-0101-01 Rev. 000 Page 6 of 260
EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I
Any Operating Mode:
R - 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 Coordinator Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)
Hot Conditions:
S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - NCS Activity 5 - NCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:
C - Cold Shutdown / Refueling System 1 - NCS Level Malfunction 2 - Loss of Essential AC Power 3 - NCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems 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 Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.
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2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, F and E) 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 (R, C, H, S, F or E)
- 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 CSD-EP-CNS-0101-01 Rev. 000 Page 8 of 260
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 Standby, 4 - Hot Shutdown, 5 -
Cold Shutdown, 6 - Refueling, D - 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:
A basis section that provides CNS-relevant information concerning the EALas well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.
CNS Basis Reference(s):
Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7) 1 Power Operation Keff > 0.99 and reactor thermal power > 5%
2 Startup Keff > 0.99 and reactor thermal power < 5%
3 Hot Standby Keff < 0.99 and average coolant temperature > 350ºF 4 Hot Shutdown Keff < 0.99 and average coolant temperature 350ºF > Tavg > 200 ºF 5 Cold Shutdown Keff < 0.99 and average coolant temperature < 200ºF 6 Refueling One or more reactor vessel head closure bolts are less than fully tensioned D Defueled Reactor vessel 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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3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Coordinator 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.
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. 4.1.12).
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 Coordinator 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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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, NCS 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 Coordinator 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 Coordinator 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 Coordinator 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. 4.1.14).
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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, whether at one unit or at two different units, 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, whether at one unit or at two different units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).
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 Coordinator 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 Coordinator, 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.
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3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
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 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 trip the reactor followed by a successful manual trip.
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 where 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. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and NCS 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.
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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 Coordinator 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.
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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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 CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents 4.1.7 CNS UFSAR Figure 1-20 Plot Plan 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control 4.1.10 PRO-NGGC-0201 NGG Procedure Writers Guide 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 CNS ISFSI Certificates of Compliance 4.1.13 CNS Emergency Plan 4.2 Implementing 4.2.1 AD-EP-ALL-0101 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to CNS EAL Comparison Matrix 4.2.3 CNS EAL Matrix CSD-EP-CNS-0101-01 Rev. 000 Page 15 of 260
5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)
Selected terms used in Initiating Condition and Emergency Action Level statements 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 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 CNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for both NAC-UMS and MAGNASTOR storage systems.
Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
As applied to CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met (ref. 4.1.9).
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)
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EPA PAGs Environmental Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires CNS to recommend protective actions for the general public to offsite planning agencies.
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.
Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
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.
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 actions that result 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.
Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.
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Hostile Action An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the 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).
Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.
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.
Maintain Take appropriate action to hold the value of an identified parameter within specified limits.
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Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.
Owner Controlled Area Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business. (ref. 4.1.13).
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 and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan (ref. 4.1.7).
NCS Intact The NCS should be considered intact when the NCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).
Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.
Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits CSD-EP-CNS-0101-01 Rev. 000 Page 19 of 260
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.
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.
Site Boundary Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents (ref. 4.1.6).
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.
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Unusual Event 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.
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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5.2 Abbreviations/Acronyms
°F ....................................................................................................... Degrees Fahrenheit
° ........................................................................................................................... Degrees AC ....................................................................................................... Alternating Current AP .................................................................................... Abnormal Operating Procedure ATWS ..................................................................... Anticipated Transient Without Scram CA ......................................................................................................Auxiliary Feedwater CNS ........................................................................................... Catawba Nuclear Station CDE ...................................................................................... Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CSFST ...................................................................... Critical Safety Function Status Tree DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current EAL ............................................................................................. Emergency Action Level ECCS............................................................................ Emergency Core Cooling System EC ............................................................................................... Emergency Coordinator ECL ................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ...............................................................................Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency ERG ................................................................................Emergency Response Guideline EPIP ................................................................Emergency Plan Implementing Procedure ESF ........................................................................................ Engineered Safety Feature FAA ................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FEMA............................................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency 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 CSD-EP-CNS-0101-01 Rev. 000 Page 22 of 260
LOCA ......................................................................................... Loss of Coolant Accident LWR .................................................................................................. Light Water Reactor MPC.............................................................................................. Multi-Purpose Canister MSIV ...................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MW .................................................................................................................... Megawatt NCS ............................................................................................ Reactor Coolant System NEI .............................................................................................. Nuclear Energy Institute NESP ................................................................... National Environmental Studies Project NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event OBE ...................................................................................... Operating Basis Earthquake OCA .............................................................................................. Owner Controlled Area ODCM........................................................................... Off-site Dose Calculation Manual ORO ................................................................................. Offsite Response Organization PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PSIG ............................................................................... Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen Rem, rem, REM ....................................................................... Roentgen Equivalent Man RPS ........................................................................................ Reactor Protection System RV .............................................................................................................Reactor Vessel RVLIS ................................................................. Reactor Vessel Level Indicating System SAR ............................................................................................... Safety Analysis Report SBGTS ......................................................................... Stand-By Gas Treatment System SBO ......................................................................................................... Station Blackout SCBA ...................................................................... Self-Contained Breathing Apparatus SG ......................................................................................................... Steam Generator CSD-EP-CNS-0101-01 Rev. 000 Page 23 of 260
SI .............................................................................................................. Safety Injection SLC ................................................................................ Selected Licensee Commitment SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SSF .............................................................................................. Safe Shutdown Facility TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center WOG .................................................................................. Westinghouse Owners Group CSD-EP-CNS-0101-01 Rev. 000 Page 24 of 260
6.0 CNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a CNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the CNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
CNS NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CSD-EP-CNS-0101-01 Rev. 000 Page 25 of 260
CNS 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 3 CG1.1 CG1 2 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 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 CSD-EP-CNS-0101-01 Rev. 000 Page 26 of 260
CNS NEI 99-01 Rev. 6 Example EAL IC EAL 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 SU8.1 SU7 1,2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 CSD-EP-CNS-0101-01 Rev. 000 Page 27 of 260
CNS NEI 99-01 Rev. 6 Example EAL IC EAL SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 CSD-EP-CNS-0101-01 Rev. 000 Page 28 of 260
7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis 7.3 Attachment 3, Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases CSD-EP-CNS-0101-01 Rev. 000 Page 29 of 260
ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 30 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity > 2 times the SLC/TS limits for 60 minutes or longer EAL:
RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for 60 min.
(Notes 1, 2, 3)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1/2EMF36L ---- ---- 4.18E+6 cpm 5.75E+3 cpm Gaseous Unit Vent Noble Gas High 1/2EMF36H 2.21E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ----
Liquid Waste Effluent Line 0EMF49L ---- ---- ---- 4.50E+6 cpm Liquid Monitor Tank Discharge 0EMF57L ---- ---- ---- 4.97E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
The column UE gaseous and liquid release values in Table R-1 represent two times the appropriate SLC and Technical Specification release rate and concentration limits associated with the specified monitors (ref. 2, 3, 4, 7).
CSD-EP-CNS-0101-01 Rev. 000 Page 31 of 260
ATTACHMENT 1 EAL Bases Gaseous Releases Instrumentation that may be used to assess this EAL is listed below (ref. 1, 5):
- Unit Vent Noble Gas Low Range - 1/2EMF36L has a range of 101 - 107 cpm Liquid Releases Instrumentation that may be used to assess this EAL is listed below (ref. 1, 6):
- Liquid Waste Effluent Line Monitor - 0EMF49L (batch release) has a range of 101 - 107 cpm
- Monitor Tank Discharge Monitor - 0EMF57L has a range of 101 - 107 cpm This IC addresses a potential decrease 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.
Escalation of the emergency classification level would be via IC RA1.
CSD-EP-CNS-0101-01 Rev. 000 Page 32 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 2. CNS-SLC 16.11-1 Liquid Effluents
- 3. CNS-SLC 16.11-6 Gaseous Effluents
- 5. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment
- 6. UFSAR Table 11-19 Liquid Process Radiation Monitoring Equipment
- 7. Technical Specifications Section 5.5.5 Radioactive Effluent Controls Program
- 8. NEI 99-01 AU1 CSD-EP-CNS-0101-01 Rev. 000 Page 33 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the SLC/TC limits for 60 minutes or longer.
EAL:
RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate
> 2 x SLC/TC limits for 60 min. (Notes 1, 2)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 decrease 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 34 of 260
ATTACHMENT 1 EAL 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 RA1.
CNS Basis Reference(s):
- 2. CNS-SLC 16.11-1 Liquid Effluents
- 3. CNS-SLC 16.11-6 Gaseous Effluents
- 4. Technical Specifications Section 5.5.5 Radioactive Effluent Controls Program
- 5. AD-RP-ALL-2003 Investigation of Unusual Radiological Occurences
- 6. NEI 99-01 AU1 CSD-EP-CNS-0101-01 Rev. 000 Page 35 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA1.1 Alert Reading on any Table R-1 effluent radiation monitor > column "ALERT" for 15 min.
(Notes 1, 2, 3, 4)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1/2EMF36L ---- ---- 4.18E+6 cpm 5.75E+3 cpm Gaseous Unit Vent Noble Gas High 1/2EMF36H 2.21E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ----
Liquid Waste Effluent Line 0EMF49L ---- ---- ---- 4.50E+6 cpm Liquid Monitor Tank Discharge 0EMF57L ---- ---- ---- 4.97E+5 cpm Mode Applicability:
All CSD-EP-CNS-0101-01 Rev. 000 Page 36 of 260
ATTACHMENT 1 EAL Bases Definition(s):
None Basis:
This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
- 10 mRem TEDE
- 50 mRem CDE Thyroid The column ALERT gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 3, 4).
Instrumentation that may be used to assess this EAL is listed below (ref. 1, 2):
- Unit Vent Noble Gas High Range - EMF36H has a range of 101 - 106 cpm 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 RS1.
CSD-EP-CNS-0101-01 Rev. 000 Page 37 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 2. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment
- 5. NEI 99-01 AA1 CSD-EP-CNS-0101-01 Rev. 000 Page 38 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA1.2 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 RA1.1, RS1.1 and RG1.1 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2)
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 39 of 260
ATTACHMENT 1 EAL Bases 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 RS1.
CNS Basis Reference(s):
- 1. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment
- 2. NEI 99-01 AA1 CSD-EP-CNS-0101-01 Rev. 000 Page 40 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA1.3 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments based on liquid releases are performed per HP/0/B/1009/014, Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material (ref. 3),
and Offsite Dose Calculation Manual (ref. 1).
This EAL addresses a release of 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 CSD-EP-CNS-0101-01 Rev. 000 Page 41 of 260
ATTACHMENT 1 EAL Bases 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 RS1.
CNS Basis Reference(s):
- 2. NEI 99-01 AA1
- 3. HP/0/B/1009/014 Radiation Protection Actions Following an Uncontrolled Release of Liquid Radioactive Material CSD-EP-CNS-0101-01 Rev. 000 Page 42 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA1.4 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2), provides guidance for emergency or post-accident radiological environmental monitoring.
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).
CSD-EP-CNS-0101-01 Rev. 000 Page 43 of 260
ATTACHMENT 1 EAL Bases 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 RS1.
CNS Basis Reference(s):
- 1. AD-EP-ALL-0203 Field Monitoring During Declared Emergency
- 2. AD-EP-CNS-0203 CNS Site Specific Field Monitoring
- 3. NEI 99-01 AA1 CSD-EP-CNS-0101-01 Rev. 000 Page 44 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RS1.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for 15 min.
(Notes 1, 2, 3, 4)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1/2EMF36L ---- ---- 4.18E+6 cpm 5.75E+3 cpm Gaseous Unit Vent Noble Gas High 1/2EMF36H 2.21E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ----
Liquid Waste Effluent Line 0EMF49L ---- ---- ---- 4.50E+6 cpm Liquid Monitor Tank Discharge 0EMF57L ---- ---- ---- 4.97E+5 cpm Mode Applicability:
All Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 45 of 260
ATTACHMENT 1 EAL Bases Basis:
This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
- 100 mRem TEDE
- 500 mRem CDE Thyroid The column SAE gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1, 2).
Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas High Range Monitor - EMF36H and has a range of 101 - 106 cpm (ref 3, 4).
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 RG1.
CNS Basis Reference(s):
- 4. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment
- 5. NEI 99-01 AS1 CSD-EP-CNS-0101-01 Rev. 000 Page 46 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RS1.2 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 RA1.1, RS1.1 and RG1.1 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2)
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 47 of 260
ATTACHMENT 1 EAL Bases 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 RG1.
CNS Basis Reference(s):
- 2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment
- 3. NEI 99-01 AS1 CSD-EP-CNS-0101-01 Rev. 000 Page 48 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RS1.3 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2) of CNS provides guidance for emergency or post-accident radiological environmental monitoring.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 49 of 260
ATTACHMENT 1 EAL Bases 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 RG1.
CNS Basis Reference(s):
- 1. AD-EP-ALL-0203 Field Monitoring During Declared Emergency
- 2. AD-EP-CNS-0203 CNS Site Specific Field Monitoring
- 3. NEI 99-01 AS1 CSD-EP-CNS-0101-01 Rev. 000 Page 50 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for 15 min.
(Notes 1, 2, 3, 4)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1/2EMF36L ---- ---- 4.18E+6 cpm 5.75E+3 cpm Gaseous Unit Vent Noble Gas High 1/2EMF36H 2.21E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ----
Liquid Waste Effluent Line 0EMF49L ---- ---- ---- 4.50E+6 cpm Liquid Monitor Tank Discharge 0EMF57L ---- ---- ---- 4.97E+5 cpm Mode Applicability:
All Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 51 of 260
ATTACHMENT 1 EAL Bases Basis:
This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
- 1000 mRem TEDE
- 5000 mRem CDE Thyroid The column GE gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1, 2).
Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas High Range Monitor - EMF36H and has a range of 101 - 106 cpm (ref 3, 4).
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.
CNS Basis Reference(s):
- 4. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment
- 5. NEI 99-01 AG1 CSD-EP-CNS-0101-01 Rev. 000 Page 52 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RG1.2 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 RA1.1, RS1.1 and RG1.1 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 2)
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 53 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment
- 3. NEI 99-01 AG1 CSD-EP-CNS-0101-01 Rev. 000 Page 54 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RG1.3 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
AD-EP-ALL-0203, Field Monitoring During Declared Emergency (ref. 1) and AD-EP-CNS-0203, CNS Site Specific Field Monitoring (ref. 2) provides guidance for emergency or post-accident radiological environmental monitoring.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 55 of 260
ATTACHMENT 1 EAL Bases 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.
CNS Basis Reference(s):
- 1. AD-EP-ALL-0203 Field Monitoring During Declared Emergency
- 2. AD-EP-CNS-0203 CNS Site Specific Field Monitoring
- 3. NEI 99-01 AG1 CSD-EP-CNS-0101-01 Rev. 000 Page 56 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:
RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
- 1EMF15 (2EMF4) Spent Fuel Building Refueling Bridge
- 1EMF17 (2EMF2) Reactor Building Refueling Bridge 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-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.
Basis:
The spent fuel pool low water level alarm setpoint is actuated by 1(2)KFPS5120 at a setpoint of 39 (ref. 1). Water level restoration instructions are performed in accordance with AOPs (ref.
2, 3).
The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 2, 3). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING CAVITY level are not classifiable under this EAL.
When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool.
CSD-EP-CNS-0101-01 Rev. 000 Page 57 of 260
ATTACHMENT 1 EAL Bases This IC addresses a decrease 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 decrease 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 the water level may also cause an increase 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 increase 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 RA2.
CNS Basis Reference(s):
- 1. OP/1(2)/B/6100/010N E/2 Spent Fuel Pool Level Hi/Lo
- 2. AP/1(2)/A/5500/026 Loss of Refueling Canal Level
- 3. AP/1(2)/A/5500/041 Loss of Spent Fuel Cooling or Level
- 4. NEI 99-01 AU2 CSD-EP-CNS-0101-01 Rev. 000 Page 58 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All Definition(s):
REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal 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 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.
This EAL escalates from RU2.1 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.
While an area radiation monitor could detect an increase 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 Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RS1.
CSD-EP-CNS-0101-01 Rev. 000 Page 59 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/026 Loss of Refueling Canal Level
- 2. AP/1(2)/A/5500/041 Loss of Spent Fuel Cooling or Level
- 3. NEI 99-01 AA2 CSD-EP-CNS-0101-01 Rev. 000 Page 60 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND A Trip 2 radiation alarm on any of the following radiation monitor indications:
- 1EMF15 (2EMF4) Spent Fuel Building Refueling Bridge
- 1EMF17 (2EMF2) Reactor Building Refueling Bridge
- 1EMF42 (2EMF42) Spent Fuel Pool Ventilation
- 1EMF39L (2EMF39L) Containment Noble Gas Mode Applicability:
All Definition(s):
None Basis:
The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel (ref. 1).
The Trip 2 alarm setpoints for the radiation monitors are set to be indicative of significant increases in area and/or airborne radiation (ref. 2).
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 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.
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 EAL EU1.1.
CSD-EP-CNS-0101-01 Rev. 000 Page 61 of 260
ATTACHMENT 1 EAL Bases Escalation of the emergency would be based on either Recognition Category R or C ICs.
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 RS1.
CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/025 Damaged Spent Fuel
- 2. HP/0/B/1000/010 Determination of Radiation Monitor Setpoints
- 3. NEI 99-01 AA2 CSD-EP-CNS-0101-01 Rev. 000 Page 62 of 260
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3 Alert Lowering of spent fuel pool level to 24.5 ft. (Level 2) on 1(2)KFP5780 or 1(2)NVP8790 Mode Applicability:
All Definition(s):
None Basis:
Post-Fukushima order EA-12-051 (ref.1) 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) and SFP level at the top of the fuel racks (Level 3).
SFP level indicators 1(2)KFP5780 (radar) or 1(2)NVP8790 (pressure) located on the back of 1(2)MC7 provide continuous wide range SFP level indication to the top of the spent fuel racks (ref. 2).
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 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.
Escalation of the emergency would be based on either Recognition Category R or C ICs.
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 RS1.
CSD-EP-CNS-0101-01 Rev. 000 Page 63 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:
RS2.1 Site Area Emergency Lowering of spent fuel pool level to 14.5 ft. (Level 3) on 1(2)KFP5780 or 1(2)NVP8790 Mode Applicability:
All Definition(s):
None Basis:
Post-Fukushima order EA-12-051 (ref.1) 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) and SFP level at the top of the fuel racks (Level 3).
SFP level indicators 1(2)KFP5780 (radar) or 1(2)NVP8790 (pressure) located on the back of 1(2)MC7 provide continuous wide range SFP level indication to the top of the spent fuel racks (ref. 2).
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 RG2.
CNS Basis Reference(s):
- 1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
- 2. EC109413
- 3. NEI 99-01 AS2 CSD-EP-CNS-0101-01 Rev. 000 Page 65 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 14.5 ft. (Level 3) on 1(2)KFP5780 or 1(2)NVP8790 for > 60 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
None Basis:
Post-Fukushima order EA-12-051 (ref.1) 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) and SFP level at the top of the fuel racks (Level 3).
SFP level indicators 1(2)KFP5780 (radar) or 1(2)NVP8790 (pressure) located on the back of 1(2)MC7 provide continuous wide range SFP level indication to the top of the spent fuel racks (ref. 2).
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.
CNS Basis Reference(s):
- 1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
- 2. EC109413
- 3. NEI 99-01 AG2 CSD-EP-CNS-0101-01 Rev. 000 Page 66 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas:
Control Room (EMF12)
OR 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 and the Central Alarm Station (CAS).
EMF Channel 12 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations.
There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the CAS.
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 Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
CSD-EP-CNS-0101-01 Rev. 000 Page 67 of 260
ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
CNS Basis Reference(s):
- 1. OP/1(2)/B/6100/010Z C/2 Control Room
- 2. NEI 99-01 AA3 CSD-EP-CNS-0101-01 Rev. 000 Page 68 of 260
ATTACHMENT 1 EAL Bases Category: R - 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:
RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (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 R-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Mode Rm 478 (1EMXA) Rm 469 (2EMXA) 4 Rm 496 (1ETA) Rm 486 (2ETA) 4 Auxiliary 577 Rm 496 (1EMXS) Rm 486 (2EMXS) 4 AB-577', JJ-57 (1MXK) AB-577', JJ-57 (2MXK) 4 Rm 330 (1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1ETB) Rm 362 (2ETB) 4 Rm 372 (1EMXD) Rm 362 (2EMXD) 4 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).
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 69 of 260
ATTACHMENT 1 EAL Bases Basis:
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 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).
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 Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For RA3.2, 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 increased 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 increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The increased 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 action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.
CSD-EP-CNS-0101-01 Rev. 000 Page 70 of 260
ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
CNS Basis Reference(s):
- 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
- 2. NEI 99-01 AA3 CSD-EP-CNS-0101-01 Rev. 000 Page 71 of 260
ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (NCS 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 NCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, D - Defueled).
The events of this category pertain to the following subcategories:
- 1. NCS Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
- 2. Loss of Essential AC Power Loss of essential 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 4160 VAC emergency buses.
- 3. NCS Temperature Uncontrolled or inadvertent temperature or pressure increases 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 125 VDC vital buses.
CSD-EP-CNS-0101-01 Rev. 000 Page 72 of 260
ATTACHMENT 1 EAL 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 73 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: UNPLANNED loss of NCS inventory for 15 minutes or longer EAL:
CU1.1 Unusual Event UNPLANNED loss of NCS inventory results in NCS water level less than a required lower limit for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - 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:
NCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
With the plant in Cold Shutdown, NCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1). However, if NCS level is being controlled below the pressurizer low level 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 NCS that is the concern.
With the plant in Refueling mode, NCS water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).
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 NCS 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 74 of 260
ATTACHMENT 1 EAL Bases Refueling evolutions that decrease NCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing 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 NCS 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 NCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/FR-I.2 Response to Low Pressurizer Level
- 2. CNS Technical Specifications Section 3.9.6 Refueling Cavity Water Level
- 3. NEI 99-01 CU1 CSD-EP-CNS-0101-01 Rev. 000 Page 75 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: UNPLANNED loss of NCS inventory EAL:
CU1.2 Unusual Event NCS water level cannot be monitored AND EITHER
- UNPLANNED increase in Containment Floor & Equipment Sump or Incore Sump (alarm) due to a loss of NCS inventory
- Visual observation of UNISOLABLE NCS leakage Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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 NCS will normally be intact and standard NCS level monitoring means are available. NCS level in the Refueling mode is normally monitored using the sight glass.
In this EAL, all water level indication is unavailable and the NCS inventory loss must be detected by indirect leakage indications (ref. 1). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of NCS leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory.
The Incore Sump level cannot be monitored in the CR but alarms on high level.
CSD-EP-CNS-0101-01 Rev. 000 Page 76 of 260
ATTACHMENT 1 EAL Bases 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 decrease NCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing 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. 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 NCS.
Continued loss of NCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/010 Reactor Coolant Leak
- 2. NEI 99-01 CU1 CSD-EP-CNS-0101-01 Rev. 000 Page 77 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory EAL:
CA1.1 Alert UNPLANNED loss of NCS inventory as indicated by NCS water level < 6.5% (wide range)
Mode Applicability:
5 - Cold Shutdown, 6 - 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:
6.5% wide range NCS level indication is the lowest level to assure adequate net positive suction head and prevent ND pump cavitation for all flow rates (ref. 1).
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 NCS water level below 6.5% indicates that operator actions have not been successful in restoring and maintaining NCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, this EAL is concerned with the loss of NCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in NCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
If NCS water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
CSD-EP-CNS-0101-01 Rev. 000 Page 78 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. OP/1(2)/A/6150/006 Draining the Reactor Coolant System
- 2. NEI 99-01 CA1 CSD-EP-CNS-0101-01 Rev. 000 Page 79 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory EAL:
CA1.2 Alert NCS water level cannot be monitored for 15 min. (Note 1)
AND EITHER
- UNPLANNED increase in Containment Floor & Equipment Sump or Incore Sump (alarm) due to a loss of NCS inventory
- Visual observation of UNISOLABLE NCS leakage Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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 NCS will normally be intact and standard NCS monitoring means are available. In the Refuel mode, the NCS is not intact and NCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all NCS water level indication would be unavailable for greater than 15 minutes, and the NCS inventory loss must be detected by indirect leakage indications (ref. 1). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory.
CSD-EP-CNS-0101-01 Rev. 000 Page 80 of 260
ATTACHMENT 1 EAL Bases The Incore Sump level cannot be monitored in the CR but alarms on high level.
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 NCS 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 NCS.
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 NCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/010 Reactor Coolant Leak
- 2. NEI 99-01 CA1 CSD-EP-CNS-0101-01 Rev. 000 Page 81 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory affecting core decay heat removal capability EAL:
CS1.1 Site Area Emergency NCS water level cannot be monitored for 30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
- UNPLANNED increase in Containment Floor & Equipment Sump or Incore Sump (alarm) due to a loss of NCS inventory
- Visual observation of UNISOLABLE NCS leakage
- Reactor Building Refueling Bridge Monitor 1EMF17 (2EMF2) reading > 9,000 mR/hr
- Erratic Source Range or Gamma Metric Monitor indication Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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:
The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop. Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS in areas outside the containment that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 82 of 260
ATTACHMENT 1 EAL Bases The Incore Sump level cannot be monitored in the CR but alarms on high level.
In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. 1EMF17 (2EMF2), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 9,000 mR/hr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
This IC addresses a significant and prolonged loss of reactor vessel/NCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a NCS 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 NCS level cannot be restored, fuel damage is probable.
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 NCS 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 NCS.
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 RG1 CSD-EP-CNS-0101-01 Rev. 000 Page 83 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. OP/1(2)/A/6150/006 Draining the Reactor Coolant System
- 2. AP/1(2)/A/5500/010 Reactor Coolant Leak
- 3. NEI 99-01 CS1 CSD-EP-CNS-0101-01 Rev. 000 Page 84 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory affecting fuel clad integrity with containment challenged EAL:
CG1.1 General Emergency NCS level cannot be monitored for 30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
- UNPLANNED increase in Containment Floor & Equipment Sump or Incore Sump (alarm) due to a loss of NCS inventory
- Visual observation of UNISOLABLE NCS leakage
- Reactor Building Refueling Bridge Monitor 1EMF17 (2EMF2) reading > 9,000 mR/hr
- Erratic Source Range or Gamma Metric Monitor indication AND Any Containment Challenge indication, Table C-1 Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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-1 Containment Challenge Indications
- CONTAINMENT CLOSURE not established (Note 6)
- Containment hydrogen concentration > 6%
- UNPLANNED rise in containment pressure Mode Applicability:
5 - Cold Shutdown, 6 - Refueling CSD-EP-CNS-0101-01 Rev. 000 Page 85 of 260
ATTACHMENT 1 EAL Bases Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
As applied to CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met.
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:
The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop. Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS in areas outside the containment that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 2).
The Incore Sump level cannot be monitored in the CR but alarms on high level.
In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. 1EMF17 (2EMF2), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 9,000 mR/hr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
CSD-EP-CNS-0101-01 Rev. 000 Page 86 of 260
ATTACHMENT 1 EAL Bases Three conditions are associated with a challenge to containment integrity:
- CONTAINMENT CLOSURE is not established (Ref. 3).
- 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 an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the containment atmosphere is greater than 6% by volume in the presence of oxygen (>5%).
- Any unplanned increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned containment pressure increases indicates containment closure cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
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 NCS 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 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 a 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 87 of 260
ATTACHMENT 1 EAL Bases 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 NCS 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 NCS.
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.
CNS Basis Reference(s):
- 1. OP/1(2)/A/6150/006 Draining the Reactor Coolant System
- 2. AP/1(2)/A/5500/010 Reactor Coolant Leak
- 3. OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode. Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control
- 4. NEI 99-01 CG1 CSD-EP-CNS-0101-01 Rev. 000 Page 88 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
CU2.1 Unusual Event AC power capability, Table C-2, to essential 4160V buses 1(2)ETA and 1(2)ETB 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-2 AC Power Sources Offsite:
- ATC (Train A)
- SATA (Train A) (if already aligned)
- ATD (Train B)
- SATB (Train B) (if already aligned)
Onsite:
- D/G A (Train A)
- D/G B (Train B)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled CSD-EP-CNS-0101-01 Rev. 000 Page 89 of 260
ATTACHMENT 1 EAL 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.
Basis:
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
This cold condition EAL is equivalent to the hot condition EAL SA1.1.
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 increased 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 90 of 260
ATTACHMENT 1 EAL Bases An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency 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.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power
- 3. NEI 99-01 CU2 CSD-EP-CNS-0101-01 Rev. 000 Page 91 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer EAL:
CA2.1 Alert Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 92 of 260
ATTACHMENT 1 EAL Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an essential 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 RS1.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power
- 3. ECA-0.0 EP/1(2)/5000/ECA-0.0 Loss of All AC Power
- 4. NEI 99-01 CA2 CSD-EP-CNS-0101-01 Rev. 000 Page 93 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:
CU3.1 Unusual Event UNPLANNED increase in NCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability:
5 - Cold Shutdown, 6 - 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:
Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
In the absence of reliable NCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should NCS level indication be subsequently lost.
This IC addresses an UNPLANNED increase in NCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
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 NCS 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.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 94 of 260
ATTACHMENT 1 EAL Bases 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 increase in reactor coolant temperature depending on the time after shutdown.
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.
CNS Basis Reference(s):
- 1. CNS Technical Specifications Table 1.1-1
- 3. AP/1(2)/A/5500/019 Loss of Residual Heat Removal System
- 4. NEI 99-01 CU3 CSD-EP-CNS-0101-01 Rev. 000 Page 95 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:
CU3.2 Unusual Event Loss of all NCS temperature and NCS level indication for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
None Basis:
Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
NCS water level is normally monitored using various instruments including NC System narrow range and wide range monitors, RVLIS, NC System sightglass, tygon tube and Pressurizer level instruments (ref. 4).
This EAL addresses the inability to determine NCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor NCS 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.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
CSD-EP-CNS-0101-01 Rev. 000 Page 96 of 260
ATTACHMENT 1 EAL Bases Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
CNS Basis Reference(s):
- 1. CNS Technical Specifications Table 1.1-1
- 3. AP/1(2)/A/5500/019 Loss of Residual Heat Removal System
- 4. OP/1(2)/A/6150/006 Draining the Reactor Coolant System
- 5. NEI 99-01 CU3 CSD-EP-CNS-0101-01 Rev. 000 Page 97 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED increase in NCS temperature to > 200°F for > Table C-3 duration (Notes 1, 9)
OR UNPLANNED NCS pressure increase > 10 psig due to a loss of NCS cooling (this does not apply during water-solid plant conditions)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 9: In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the NCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6.
Table C-3: NCS Heat-up Duration Thresholds Containment Closure NCS Status Heat-up Duration Status Intact (but not reduced N/A 60 min.*
inventory)
Not intact established 20 min.*
OR not established 0 min.
At reduced inventory
- If an NCS heat removal system is in operation within this time frame and NCS temperature is being reduced, the EAL is not applicable.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
CSD-EP-CNS-0101-01 Rev. 000 Page 98 of 260
ATTACHMENT 1 EAL Bases As applied to CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met.
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 NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
A 10 psig RPV pressure increase can be read on various instruments such as NCPT5141 C-Loop N/R, 0 - 800 psig and Point #4 on SMCR5810 (CR chart recorder, 0 - 600 psi). (ref. 4, 5).
In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the NCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6.
RCS reduced inventory condition exists when NCS level is 16% (ref. 7).
This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the NCS 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 NCS Heat-up Duration Thresholds table addresses an increase in NCS temperature when CONTAINMENT CLOSURE is established but the NCS is not intact, or NCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The NCS Heat-up Duration Thresholds table also addresses an increase in NCS temperature with the NCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact NCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
CSD-EP-CNS-0101-01 Rev. 000 Page 99 of 260
ATTACHMENT 1 EAL Bases Finally, in the case where there is an increase in NCS temperature, the NCS is not intact or is at reduced inventory, 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 NCS pressure increase threshold provides a pressure-based indication of NCS heat-up in the absence of NCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC CS1 or RS1.
CNS Basis Reference(s):
- 1. CNS Technical Specifications Table 1.1-1
- 3. AP/1(2)/A/5500/019 Loss of Residual Heat Removal System
- 4. IP/1(2)/B/3121/011A
- 5. IP/1(2)/A/3122/055A
- 6. OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode. Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control
- 7. OP/1(2)/A/6150/006 Draining the Reactor Coolant System
- 8. NEI 99-01 CA3 CSD-EP-CNS-0101-01 Rev. 000 Page 100 of 260
ATTACHMENT 1 EAL 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
< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
Four 125 VDC distribution centers are provided for the 125VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDC, EDB and EDD), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-120VAC static inverter (ref. 1).
The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1, 2).
This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.
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 increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.
CSD-EP-CNS-0101-01 Rev. 000 Page 101 of 260
ATTACHMENT 1 EAL Bases 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.
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 R.
CNS Basis Reference(s):
- 2. AP/1(2)/A/5500/029 Loss of Vital or Aux Control Power
- 3. NEI 99-01 CU4 CSD-EP-CNS-0101-01 Rev. 000 Page 102 of 260
ATTACHMENT 1 EAL 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-4 onsite communication methods OR Loss of all Table C-4 ORO communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite ORO NRC Public Address X Internal Telephones X Onsite Radios X DEMNET X Commercial Telephones X X Satellite Phones X X Cellular Phones X X NRC Emergency Telephone System (ETS) X X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 103 of 260
ATTACHMENT 1 EAL Bases Basis:
Onsite/offsite communications include one or more of the systems listed in Table C-4 (ref. 1).
Public Address System The Catawba Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
Internal Telephone System The Catawba Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code.
On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EOF.
DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office.
Satellite Phones Portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major external event. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources.
Cellular Phones Cellular phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.
CSD-EP-CNS-0101-01 Rev. 000 Page 104 of 260
ATTACHMENT 1 EAL Bases NRC Emergency Telephone System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Catawba Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
This EAL is the cold condition equivalent of the hot condition EAL SU7.1.
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 OROs 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 OROs of an emergency declaration. The OROs referred to here are the State, York, Gaston and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
CNS Basis Reference(s):
- 1. CNS Emergency Plan Section F Emergency Communications
- 2. NEI 99-01 CU5 CSD-EP-CNS-0101-01 Rev. 000 Page 105 of 260
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
CA6.1 Alert The occurrence of any Table C-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 one of the following:
Event damage has caused indications of degraded performance to a 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 (Note 11, 12)
Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
Note 12: 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-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 Mode Applicability:
6 - Cold Shutdown, 6 - Refueling CSD-EP-CNS-0101-01 Rev. 000 Page 106 of 260
ATTACHMENT 1 EAL Bases 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:
- The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
- Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).
The minimum external access elevation for the Auxiliary, Turbine and Service Buildings is 594.0 ft MSL (ref. 1, 3).
CSD-EP-CNS-0101-01 Rev. 000 Page 107 of 260
ATTACHMENT 1 EAL Bases
- Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 95 mph. (ref. 4).
- Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5).
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 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 for criteria in the first condition of this EAL; 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 address 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.
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.
An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under CA6, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met.
If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
CSD-EP-CNS-0101-01 Rev. 000 Page 108 of 260
ATTACHMENT 1 EAL Bases An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under CA6, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC RS1.
CNS Basis Reference(s):
- 1. RP/0/A/5000/007 Natural Disaster and Earthquake
- 2. AP/0/A/5500/030 Plant Flooding
- 3. UFSAR Section 3.4 Water Level (Flood) Design
- 4. Updated FSAR Section 3.3.1 Wind Loadings
- 5. AP/0/A/5500/045 Plant Fire
- 6. NEI 99-01 CA6 CSD-EP-CNS-0101-01 Rev. 000 Page 109 of 260
ATTACHMENT 1 EAL 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 Technology 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 site 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 110 of 260
ATTACHMENT 1 EAL Bases
- 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 Coordinator 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 on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator judgment.
CSD-EP-CNS-0101-01 Rev. 000 Page 111 of 260
ATTACHMENT 1 EAL 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 the Security Shift Supervision Mode Applicability:
All Definition(s):
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.
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
Basis:
This EAL is based on the Duke Energy Physical Security Plan for CNS (ref. 1).
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 112 of 260
ATTACHMENT 1 EAL Bases Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
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.
This 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.
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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HA1.
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. AP/0/A/5500/046 Hostile Aircraft Activity
- 3. RP/0/B/5000/026 Site Response to a Security Threat
- 4. AP/0/A/5500/048 Extensive Damage Mitigation
- 5. NEI 99-01 HU1 CSD-EP-CNS-0101-01 Rev. 000 Page 113 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:
HU1.2 Unusual Event Notification of a credible security threat directed at the site Mode Applicability:
All Definition(s):
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 EAL is based on the Duke Energy Physical Security Plan for CNS (ref. 1).
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 (ref. 2, 3, 4).
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.
This threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the CNS Security Contingency Plan (ref. 1).
CSD-EP-CNS-0101-01 Rev. 000 Page 114 of 260
ATTACHMENT 1 EAL Bases 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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HA1.
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. AP/0/A/5500/046 Hostile Aircraft Activity
- 3. RP/0/B/5000/026 Site Response to a Security Threat
- 4. AP/0/A/5500/048 Extensive Damage Mitigation
- 5. NEI 99-01 HU1 CSD-EP-CNS-0101-01 Rev. 000 Page 115 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:
HU1.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
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 EAL is based on the Duke Energy Physical Security Plan for CNS (ref. 1).
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 (ref. 2, 3, 4).
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.
This 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 the CNS Security Contingency Plan (ref. 1).
CSD-EP-CNS-0101-01 Rev. 000 Page 116 of 260
ATTACHMENT 1 EAL Bases 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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HA1.
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. AP/0/A/5500/046 Hostile Aircraft Activity
- 3. RP/0/B/5000/026 Site Response to a Security Threat
- 4. AP/0/A/5500/048 Extensive Damage Mitigation
- 5. NEI 99-01 HU1 CSD-EP-CNS-0101-01 Rev. 000 Page 117 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat EAL:
HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the 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 the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 118 of 260
ATTACHMENT 1 EAL Bases 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 IC 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.
This threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. AP/0/A/5500/046 Hostile Aircraft Activity
- 3. RP/0/B/5000/026 Site Response to a Security Threat
- 4. AP/0/A/5500/048 Extensive Damage Mitigation
- 5. NEI 99-01 HA1 CSD-EP-CNS-0101-01 Rev. 000 Page 119 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat EAL:
HA1.2 Alert A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the 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 the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 120 of 260
ATTACHMENT 1 EAL Bases 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 IC 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.
This 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 site-specific security procedures.
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.
In some cases, it may not be readily apparent if an aircraft impact within the 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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. AP/0/A/5500/046 Hostile Aircraft Activity
- 3. RP/0/B/5000/026 Site Response to a Security Threat
- 4. AP/0/A/5500/048 Extensive Damage Mitigation
- 5. NEI 99-01 HA1 CSD-EP-CNS-0101-01 Rev. 000 Page 121 of 260
ATTACHMENT 1 EAL 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 the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for CNS (Safeguards) information. (ref. 1)
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. 2, 3).
CSD-EP-CNS-0101-01 Rev. 000 Page 122 of 260
ATTACHMENT 1 EAL Bases 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 IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. It also 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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
- 1. Duke Energy Physical Security Plan for CNS
- 2. RP/0/B/5000/026 Site Response to a Security Threat
- 3. AP/0/A/5500/048 Extensive Damage Mitigation
- 4. NEI 99-01 HS1 CSD-EP-CNS-0101-01 Rev. 000 Page 123 of 260
ATTACHMENT 1 EAL 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 OBE EXCEEDED alarm on 1AD-4, B/8 Mode Applicability:
All Definition(s):
None Basis:
Ground motion acceleration of 0.08g horizontal or 0.0533g vertical is the Operating Basis Earthquake for CNS (ref. 1).
Five strong motion accelerographs are installed within Unit 1 structures. The seismic instrumentation system also consists of a network control center (NCC), which is used for rapid interrogation of the accelerograph data and for data transfer to a dedicated system computer for subsequent data processing and analysis. The time-history recorded at each accelerograph location can be analyzed to determine its corresponding peak acceleration values and to verify that site Operating Basis Earthquake (OBE) limits have not been exceeded. Immediate control room alarm indication of an earthquake of 0.08 g horizontal or 0.533 g vertical or greater is annunciated through the system's network control center (NCC),
following seismic trigger actuation by at least two accelerographs (ref. 2).
RP/0/A/5000/007 Natural Disaster and Earthquake provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 3)
CSD-EP-CNS-0101-01 Rev. 000 Page 124 of 260
ATTACHMENT 1 EAL 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) 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. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of CNS. Provide the analyst with the following CNS coordinates: 35º 03' 04" north latitude, 81º 04' 10" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website:
http://earthquake.usgs.gov/eqcenter/
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., performs 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 (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the 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 IC CA6 or SA9.
CNS Basis Reference(s):
- 1. Updated FSAR Section 3.1 Conformance with General Design Criteria
- 2. Updated FSAR Section 3.7.4.2 Location and Description of Instrumentation
- 3. RP/0/A/5000/007 Natural Disaster and Earthquake
- 4. Updated FSAR section 2.1.1.1 Specification of Location (Unit 1)
- 5. NEI 99-01 HU2 CSD-EP-CNS-0101-01 Rev. 000 Page 125 of 260
ATTACHMENT 1 EAL 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 and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
Response actions associated with a tornado onsite is provided in RP/0/A/5000/007 Enclosure 4.2 Tornado Warning Issued for York County or Tornado On-Site (ref. 1).
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
CNS Basis Reference(s):
- 1. RP/0/A/5000/007 Natural Disaster and Earthquake
- 2. NEI 99-01 HU3 CSD-EP-CNS-0101-01 Rev. 000 Page 126 of 260
ATTACHMENT 1 EAL 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 needed 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:
Areas susceptible to internal flooding are Turbine/Service Buildings and Auxiliary/Diesel Buildings from the following systems: Condenser Circulating Water, Fire Protection, Nuclear and Conventional Service Water and Condensate Storage (ref.1). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEM trains.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
CSD-EP-CNS-0101-01 Rev. 000 Page 127 of 260
ATTACHMENT 1 EAL Bases 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 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 R, F, S or C.
CNS Basis Reference(s):
- 1. AP/0/A/5500/030 Plant Flooding
- 2. NEI 99-01 HU3 CSD-EP-CNS-0101-01 Rev. 000 Page 128 of 260
ATTACHMENT 1 EAL 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 offsite event 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 and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
As used here, the term "offsite" is meant to be areas external to the CNS PROTECTED AREA.
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 an offsite location 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 R, F, S or C.
CNS Basis Reference(s):
- 1. NEI 99-01 HU3 CSD-EP-CNS-0101-01 Rev. 000 Page 129 of 260
ATTACHMENT 1 EAL 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):
None 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 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 R, F, S or C.
CNS Basis Reference(s):
- 1. NEI 99-01 HU3 CSD-EP-CNS-0101-01 Rev. 000 Page 130 of 260
ATTACHMENT 1 EAL 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-1 Fire Areas
- Reactor Building (Containment)
- Auxiliary Building
- Diesel Generator Rooms
- RN Pump House
- Dog Houses
- Standby Shutdown Facility (SSF)
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 131 of 260
ATTACHMENT 1 EAL Bases Basis:
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 based on CNS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection and AP/0/A/5500/045 Plant Fire. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).
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.
For EAL HU4.1 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.
CNS Basis Reference(s):
- 1. CNS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection
- 2. AP/0/A/5500/045 Plant Fire
- 3. NEI 99-01 HU4 CSD-EP-CNS-0101-01 Rev. 000 Page 132 of 260
ATTACHMENT 1 EAL 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)
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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-1 Fire Areas
- Reactor Building (Containment)
- Auxiliary Building
- Diesel Generator Rooms
- RN Pump House
- Dog Houses
- Standby Shutdown Facility (SSF)
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 133 of 260
ATTACHMENT 1 EAL Bases Basis:
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.
Table H-1 Fire Areas are based on CNS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection and AP/0/A/5500/045 Plant Fire. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).
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.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 134 of 260
ATTACHMENT 1 EAL Bases 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.
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 this EAL, 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 IC CA6 or SA9.
CNS Basis Reference(s):
- 1. CNS-1465.00-00-0006 Design Basis Specification for the Plant Fire Protection
- 2. AP/0/A/5500/045 Plant Fire
- 3. NEI 99-01 HU4 CSD-EP-CNS-0101-01 Rev. 000 Page 135 of 260
ATTACHMENT 1 EAL 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 plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
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 IC CA6 or SA9.
CNS Basis Reference(s):
- 1. NEI 99-01 HU4 CSD-EP-CNS-0101-01 Rev. 000 Page 136 of 260
ATTACHMENT 1 EAL 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 plant 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 and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
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 IC CA6 or SA9.
CNS Basis Reference(s):
- 1. NEI 99-01 HU4 CSD-EP-CNS-0101-01 Rev. 000 Page 137 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases 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 rooms or areas 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 Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Mode Rm 478 (1EMXA) Rm 469 (2EMXA) 4 Rm 496 (1ETA) Rm 486 (2ETA) 4 Auxiliary 577 Rm 496 (1EMXS) Rm 486 (2EMXS) 4 AB-577', JJ-57 (1MXK) AB-577', JJ-57 (2MXK) 4 Rm 330 (1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1ETB) Rm 362 (2ETB) 4 Rm 372 (1EMXD) Rm 362 (2EMXD) 4 Mode Applicability:
4 - 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).
CSD-EP-CNS-0101-01 Rev. 000 Page 138 of 260
ATTACHMENT 1 EAL Bases Basis:
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 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).
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 Coordinators 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).
CSD-EP-CNS-0101-01 Rev. 000 Page 139 of 260
ATTACHMENT 1 EAL Bases 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 4.
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 action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
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.
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 automatically or manually activate a fire suppression system in an area.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
NOTE: IC HA5 mode applicability has been limited to the applicable modes identified in Table H-2 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases and to IC HA5 mode applicability is required.
CNS Basis Reference(s):
- 1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
- 2. NEI 99-01 HA5 CSD-EP-CNS-0101-01 Rev. 000 Page 140 of 260
ATTACHMENT 1 EAL 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 Auxiliary Shutdown Panels or Standby Shutdown Facility Mode Applicability:
All Definition(s):
None Basis:
The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).
Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 141 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/017 Loss of Control Room
- 2. OP/1(2)/A/6100/004 Shutdown Outside the Control Room From Hot Standby to Cold Shutdown.
- 3. NEI 99-01 HA6 CSD-EP-CNS-0101-01 Rev. 000 Page 142 of 260
ATTACHMENT 1 EAL 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 Auxiliary Shutdown Panels or Standby Shutdown Facility AND Control of any of the following key safety functions is not reestablished within 15 min.
(Note 1):
Reactivity (Modes 1, 2 and 3 only)
Core Cooling NCS heat removal Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 -
Refueling Definition(s):
None Basis:
The Shift Manager determines if the Control Room is inoperable and requires evacuation.
Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 143 of 260
ATTACHMENT 1 EAL Bases The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment. The Emergency Coordinator 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).
Escalation of the emergency classification level would be via IC FG1 or CG1 CNS Basis Reference(s):
- 1. AP/1(2)/A/5500/017 Loss of Control Room
- 2. OP/1(2)/A/6100/004 Shutdown Outside the Control Room From Hot Standby to Cold Shutdown.
- 3. NEI 99-01 HS6 CSD-EP-CNS-0101-01 Rev. 000 Page 144 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:
HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Coordinator 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 145 of 260
ATTACHMENT 1 EAL Bases Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).
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 Coordinator to fall under the emergency classification level description for an Unusual Event.
CNS Basis Reference(s):
- 1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HU7 CSD-EP-CNS-0101-01 Rev. 000 Page 146 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:
HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Coordinator, 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):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).
CSD-EP-CNS-0101-01 Rev. 000 Page 147 of 260
ATTACHMENT 1 EAL Bases 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 Coordinator to fall under the emergency classification level description for an Alert.
CNS Basis Reference(s):
- 1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HA7 CSD-EP-CNS-0101-01 Rev. 000 Page 148 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:
HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator 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):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).
SITE BOUNDARY - Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
CSD-EP-CNS-0101-01 Rev. 000 Page 149 of 260
ATTACHMENT 1 EAL Bases Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).
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 Coordinator to fall under the emergency classification level description for a Site Area Emergency.
CNS Basis Reference(s):
- 1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HS7 CSD-EP-CNS-0101-01 Rev. 000 Page 150 of 260
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:
HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Coordinator 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):
HOSTILE ACTION - An act toward CNS 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 CNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 151 of 260
ATTACHMENT 1 EAL Bases Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager(SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).
Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.
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 Coordinator to fall under the emergency classification level description for a General Emergency.
CNS Basis Reference(s):
- 1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HG7 CSD-EP-CNS-0101-01 Rev. 000 Page 152 of 260
ATTACHMENT 1 EAL Bases Category S - System Malfunction EAL Group: Hot Conditions (NCS 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 Essential AC Power Loss of emergency 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 sources for 4160 V essential buses.
- 2. Loss of Vital DC Power Loss of emergency 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 vital plant 125 VDC power sources.
- 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. NCS 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 increase 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 153 of 260
ATTACHMENT 1 EAL Bases
- 5. NCS Leakage The reactor 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 NCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, NCS and containment integrity.
- 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip 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 trip 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, NCS 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. Containment Isolation Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification.
- 9. 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 154 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite AC power capability to essential buses for 15 minutes or longer EAL:
SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1 AC Power Sources Offsite:
- ATC (Train A)
- SATA (Train A) (if already aligned)
- ATD (Train B)
- SATB (Train B) (if already aligned)
Onsite:
- D/G A (Train A)
- D/G B (Train B)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 155 of 260
ATTACHMENT 1 EAL Bases Basis:
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.
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 essential 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 essential buses, whether or not the buses are powered from it.
Escalation of the emergency classification level would be via IC SA1.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power
- 3. NEI 99-01 SU1 CSD-EP-CNS-0101-01 Rev. 000 Page 156 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
SA1.1 Alert AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1(2)ETB 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1 AC Power Sources Offsite:
- ATC (Train A)
- SATA (Train A) (if already aligned)
- ATD (Train B)
- SATB (Train B) (if already aligned)
Onsite:
- D/G A (Train A)
- D/G B (Train B)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown CSD-EP-CNS-0101-01 Rev. 000 Page 157 of 260
ATTACHMENT 1 EAL 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.
Basis:
For emergency classification purposes, capability means that an AC power source is available to the essential buses, whether or not the buses are powered from it.
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.
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.
CSD-EP-CNS-0101-01 Rev. 000 Page 158 of 260
ATTACHMENT 1 EAL Bases An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.
Escalation of the emergency classification level would be via IC SS1.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power
- 3. NEI 99-01 SA1 CSD-EP-CNS-0101-01 Rev. 000 Page 159 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to essential buses for 15 minutes or longer EAL:
SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This EAL is indicated by the loss of all offsite and onsite AC power capability (Table S-1) to 4160V essential buses ETA and ETB. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1). For emergency classification purposes, capability means that an AC power source is available to the essential buses, whether or not the buses are powered from it.
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
CSD-EP-CNS-0101-01 Rev. 000 Page 160 of 260
ATTACHMENT 1 EAL Bases Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power capability are lost.
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. 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.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power
- 3. ECA-0.0 EP/1(2)/5000/ECA-0.0 Loss of All AC Power
- 4. NEI 99-01 SS1 CSD-EP-CNS-0101-01 Rev. 000 Page 161 of 260
ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to essential buses EAL:
SG1.1 General Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1(2)ETB AND SSF fails to supply NC pump seal injection OR CA supply to SGs AND EITHER:
- Restoration of at least one essential 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)
- Core Cooling RED PATH conditions met Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 162 of 260
ATTACHMENT 1 EAL Bases Basis:
This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4160V emergency buses ETA and ETB either for greater then the CNS Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual loss of adequate core cooling.
The SSF is capable of providing the necessary functions (reactor coolant pump seal injection and auxiliary feedwater supply to the steam generators) to maintain a hot shutdown condition for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. No fission product barrier degradation would be expected if the SSF is functioning as intended.
Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. (ref. 2).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2).
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 3).
Four hours is the station blackout coping time (ref 2).
Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Coordinator judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2.
This IC addresses a prolonged loss of all power sources to AC essential 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. 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 163 of 260
ATTACHMENT 1 EAL Bases 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 essential 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 an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one essential 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.
CNS Basis Reference(s):
- 1. UFSAR Section 8.4.2 Station Blackout Duration
- 2. EP/1/A/5000/F-0 Critical Safety Function Status Tress - Core Cooling
- 3. UFSAR Section 8.0 Electric Power
- 4. AP/1(2)/A/5500/007 Loss of Normal Power
- 5. ECA-0.0 EP/1(2)/5000/ECA-0.0 Loss of All AC Power
- 6. NEI 99-01 SG1 CSD-EP-CNS-0101-01 Rev. 000 Page 164 of 260
ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all 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 capability, to essential 4160V buses 1(2)ETA and 1(2)ETB for 15 min.
AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDD, EDB and EDC for 15 min.
(Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4160V emergency buses ETA and ETB for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
CSD-EP-CNS-0101-01 Rev. 000 Page 165 of 260
ATTACHMENT 1 EAL Bases Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Safe Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event (ref. 1).
Four 125 VDC distribution centers are provided for the 125VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDD, EDB and EDC), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-120VAC static inverter (ref. 1, 3).
The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1, 3).
This IC addresses a concurrent and prolonged loss of both essential AC and Vital DC power.
A loss of all essential 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. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both essential 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.
CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power
- 2. AP/1(2)/A/5500/007 Loss of Normal Power 3 AP/1(2)/A/5500/029 Loss of Vital or Aux Control Power
- 4. ECA-0.0 EP/1(2)/5000/ECA-0.0 Loss of All AC Power
- 5. NEI 99-01 SG8 CSD-EP-CNS-0101-01 Rev. 000 Page 166 of 260
ATTACHMENT 1 EAL 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 Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDC, EDB, EDD and for 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Four 125 VDC distribution centers are provided for the 125VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDC, EDB and EDD), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-120VAC static inverter (ref. 1, 2).
The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 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.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.
CSD-EP-CNS-0101-01 Rev. 000 Page 167 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power 2 AP/1(2)/A/5500/029 Loss of Vital or Aux Control Power
- 3. NEI 99-01 SS8 CSD-EP-CNS-0101-01 Rev. 000 Page 168 of 260
ATTACHMENT 1 EAL 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-2 Safety System Parameters
- Reactor power
- NCS level
- NCS pressure
- In-core T/C temperature
- Level in at least one S/G
- Auxiliary or emergency feed flow in at least one S/G Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown 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:
SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 169 of 260
ATTACHMENT 1 EAL Bases 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 and 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, core cooling and NCS 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 reactor vessel 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 SA3.
CNS Basis Reference(s):
- 1. UFSAR Section 7.5 Safety-Related Display Instrumentation
- 2. OP/1(2)/A/6700/003 Operation With the Operator Aid Computer Out of Service
- 3. NEI 99-01 SU2 CSD-EP-CNS-0101-01 Rev. 000 Page 170 of 260
ATTACHMENT 1 EAL 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-2 Safety System Parameters
- Reactor power
- NCS level
- NCS pressure
- In-core T/C temperature
- Level in at least one S/G
- Auxiliary or emergency feed flow in at least one S/G Table S-3 Significant Transients
- Runback > 25% thermal power
- Electrical load rejection > 25%
electrical load
- Safety injection actuation CSD-EP-CNS-0101-01 Rev. 000 Page 171 of 260
ATTACHMENT 1 EAL Bases Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown 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:
SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.
The Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).
Significant transients are listed in Table S-2 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load or SI injection actuations.
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 and 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 172 of 260
ATTACHMENT 1 EAL Bases This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and NCS 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 reactor vessel 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 ICs FS1 or IC RS1.
CNS Basis Reference(s):
- 1. UFSAR Section 7.5 Safety-Related Display Instrumentation
- 2. OP/1(2)/A/6700/003 Operation with the Operator Aid Computer Out of Service
- 3. NEI 99-01 SA2 CSD-EP-CNS-0101-01 Rev. 000 Page 173 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 4 - NCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1 Unusual Event NCS activity > Technical Specification 3.4.16 limits Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Technical Specification Section 3.4.16, as modified in the Facility Operating License, limits NC System Dose Equivalent I-131 to 0.46 µCi/gm. Technical Specification Section 3.4.16 also limits NC System Dose Equivalent Xe-133 to 280 µCi/gm. (ref 1, 2).
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 ICs FA1 or the Recognition Category R ICs.
CNS Basis Reference(s):
- 1. CNS Technical Specifications section 3.4.16 RCS Specific Activity
- 2. Facility Operating License Attachment B
- 3. NEI 99-01 SU3 CSD-EP-CNS-0101-01 Rev. 000 Page 174 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - NCS Leakage Initiating Condition: NCS leakage for 15 minutes or longer EAL:
SU5.1 Unusual Event NCS unidentified or pressure boundary leakage > 10 gpm for 15 min.
OR NCS identified leakage > 25 gpm for 15 min.
OR Leakage from the NCS to a location outside containment > 25 gpm for 15 min.
(Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Identified leakage includes leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, leakage into the containment 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; or NCS leakage through a steam generator to the secondary system (ref. 1).
Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 1).
Pressure Boundary leakage is leakage (except SG leakage) through an unisolable fault in an NCS component body, pipe wall, or vessel wall (ref. 1)
NCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as NCS to the Component Cooling Water (KC), or systems that directly see NCS pressure outside containment such as Chemical & Volume Control System (NV), Nuclear Sampling system (NM) and Residual Heat Removal (ND) system (when in the shutdown cooling mode) (ref. 2)
CSD-EP-CNS-0101-01 Rev. 000 Page 175 of 260
ATTACHMENT 1 EAL Bases Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1.
This IC addresses NCS leakage which may be a precursor to a more significant event. In this case, NCS 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 NCS 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 NCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) 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 NCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Failure to isolate the leak within 15 minutes, or if known that leak cannot be isolated within 15 minutes from the start of the leak, requires immediate classification.
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
CNS Basis Reference(s):
- 1. CNS Technical Specifications Definitions section 1.1
- 2. UFSAR Section 5.2.5.2.1 Intersystem Leakage
- 3. NEI 99-01 SU4 CSD-EP-CNS-0101-01 Rev. 000 Page 176 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:
SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is success in shutting down the reactor as indicated by reactor power < 5% (Note 8)
Note 8: A manual trip 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 Definition(s):
None Basis:
The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).
CSD-EP-CNS-0101-01 Rev. 000 Page 177 of 260
ATTACHMENT 1 EAL Bases For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
Following any automatic RPS trip signal, EP/1(2)/A/5000/E-0 (ref. 2) and EP/1(2)/A/5000/FR-S.1 (ref. 3) prescribe insertion of redundant manual trip signals to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event (ref.
4).
In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power below 5%, the event escalates to the Alert under EAL SA6.1.
If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.
If a subsequent review of the trip actuation indications reveals that the automatic trip 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.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip 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.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 178 of 260
ATTACHMENT 1 EAL Bases If an initial manual reactor trip 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 trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip 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 trip). 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.
The plant response to the failure of an automatic or manual reactor trip 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
CNS Basis Reference(s):
- 1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
- 4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SU5 CSD-EP-CNS-0101-01 Rev. 000 Page 179 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:
SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is success in shutting down the reactor as indicated by reactor power < 5% (Note 8)
Note 8: A manual trip 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 Definition(s):
None Basis:
This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power < 5%). (ref. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from a manual reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3 4).
CSD-EP-CNS-0101-01 Rev. 000 Page 180 of 260
ATTACHMENT 1 EAL Bases For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
If both subsequent automatic and subsequent manual reactor trip 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 trip, the event escalates to an Alert under EAL SA6.1 This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip 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.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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 trip 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 trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip 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 trip). 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.
The plant response to the failure of an automatic or manual reactor trip 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.
CSD-EP-CNS-0101-01 Rev. 000 Page 181 of 260
ATTACHMENT 1 EAL Bases A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
CNS Basis Reference(s):
- 1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
- 4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SU5 CSD-EP-CNS-0101-01 Rev. 000 Page 182 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Automatic or manual trip 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 trip fails to shut down the reactor as indicated by reactor power
> 5%
AND Manual trip actions taken at the reactor control console (manual reactor trip switches or turbine manual trip) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8)
Note 8: A manual trip 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 Definition(s):
None Basis:
This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip 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.
For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
CSD-EP-CNS-0101-01 Rev. 000 Page 183 of 260
ATTACHMENT 1 EAL Bases 5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1).
Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip 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 RTS.
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 trip). 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 console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control console.
The plant response to the failure of an automatic or manual reactor trip 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 the core cooling or NCS 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.
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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
CSD-EP-CNS-0101-01 Rev. 000 Page 184 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
- 4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SA5 CSD-EP-CNS-0101-01 Rev. 000 Page 185 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or NCS heat removal EAL:
SS6.1 Site Area Emergency An automatic or manual trip 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:
- Core Cooling RED PATH conditions met
- Heat Sink RED PATH conditions met Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses the following:
- Any automatic reactor trip signal followed by a manual trip 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.
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 NCS barriers.
CSD-EP-CNS-0101-01 Rev. 000 Page 186 of 260
ATTACHMENT 1 EAL Bases Reactor shutdown achieved by use of EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 4).
5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1, 4).
Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2).
Indication of inability to adequately remove heat from the NCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 3).
This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip 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 NCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 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 IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG1 or FG1.
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
- 2. EP/1(2)/A/5000/F-0 Critical Safety Function Status Tress - Core Cooling
- 3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Tress - Heat Sink
- 4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SS5 CSD-EP-CNS-0101-01 Rev. 000 Page 187 of 260
ATTACHMENT 1 EAL 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 ORO communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite ORO NRC Public Address X Internal Telephones X Onsite Radios X DEMNET X Commercial Telephones X X Satellite Phones X X Cellular Phones X X NRC Emergency Telephone System (ETS) X X Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None CSD-EP-CNS-0101-01 Rev. 000 Page 188 of 260
ATTACHMENT 1 EAL Bases Basis:
Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1).
Public Address System The Catawba Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
Internal Telephone System The Catawba Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code.
On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EOF.
DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office.
Satellite Phones A portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major external event. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources.
Cellular Phones Cellular phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.
CSD-EP-CNS-0101-01 Rev. 000 Page 189 of 260
ATTACHMENT 1 EAL Bases NRC Emergency Telephone System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Catawba Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
This EAL is the hot condition equivalent of the cold condition EAL CU5.1.
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 OROs 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 OROs of an emergency declaration. The OROs referred to here are the State, York, Gaston and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
CNS Basis Reference(s):
- 1. CNS Emergency Plan Section F Emergency Communications
- 2. NEI 99-01 SU6 CSD-EP-CNS-0101-01 Rev. 000 Page 190 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control.
EAL:
SU8.1 Unusual Event EITHER:
Any penetration is not isolated within 15 min. of a VALID containment isolation signal (Note 1)
OR Containment pressure > 3 psig with < one full train of containment cooling operating per design for > 15 min. (Notes 1, 10)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown 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:
The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.
One full train of containment cooling operating per design is considered (ref. 1, 2):
- One train of Containment Air Return Fan System (VX-CARF), and
- One train of Containment Spray System (NS)
CSD-EP-CNS-0101-01 Rev. 000 Page 191 of 260
ATTACHMENT 1 EAL Bases Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the CARF with a 10-minute delay) within 15 minutes this EAL has been exceeded. At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control.
The Unusual Event threshold applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes.
The Unusual Event threshold also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the Unusual Event should be declared regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.
This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or NCS fission product barriers.
CNS Basis Reference(s):
- 2. CNS Technical Specification 3.6.6 Bases
- 4. UFSAR Section 6.2 Containment Systems
- 5. NEI 99-01 SU7 CSD-EP-CNS-0101-01 Rev. 000 Page 192 of 260
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
SA9.1 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 of the following:
Event damage has caused indications of degraded performance to a 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 (Note 11, 12)
Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
Note 12: 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 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - 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, CSD-EP-CNS-0101-01 Rev. 000 Page 193 of 260
ATTACHMENT 1 EAL Bases 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:
- The significance of seismic events are discussed under EAL HU2.1 (ref. 1).
- Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).
The minimum external access elevation for the Auxiliary, Turbine and Service Buildings is 594.0 ft MSL (ref. 1, 3).
- Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 95 mph. (ref. 4).
- Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5).
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 CSD-EP-CNS-0101-01 Rev. 000 Page 194 of 260
ATTACHMENT 1 EAL Bases 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 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 for criteria in the first condition of this EAL; 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 address 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.
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.
An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under SA9, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA9 because the two-train impact criteria that underlie the EALs and Bases would not be met.
If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under SA9, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC FS1 or RS1.
CSD-EP-CNS-0101-01 Rev. 000 Page 195 of 260
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
- 1. RP/0/A/5000/007 Natural Disaster and Earthquake
- 2. AP/0/A/5500/030 Plant Flooding
- 3. UFSAR Section 3.4 Water Level (Flood) Design
- 4. Updated FSAR Section 3.3.1 Wind Loadings
- 5. AP/0/A/5500/045 Plant Fire
- 6. NEI 99-01 SA9 CSD-EP-CNS-0101-01 Rev. 000 Page 196 of 260
ATTACHMENT 1 EAL 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 cask/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 CNS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISFSI would be applicable to EALs HU1.1, HA1.1and HS1.1 Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.
CSD-EP-CNS-0101-01 Rev. 000 Page 197 of 260
ATTACHMENT 1 EAL Bases Category: E - ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1.1 Notification of Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask> any Table E-1 dose limit.
- 240 mrem/hr (gamma) on the vertical
- 100 mrem/hr (neutron + gamma) on concrete surfaces the side (on the concrete surfaces)
- 10 mrem/hr (neutron) on the vertical
- 100 mrem/hr (neutron + gamma) on concrete surfaces the top
- 900 mrem/hr (neutron + gamma) on the
- 200 mrem/hr (neutron + gamma) at air top inlets and outlets 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 CNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for both NAC-UMS and MAGNASTOR storage systems.
Basis:
The CNS ISFSI utilizes two designs for dry spent fuel storage:
- The NAC-UMS dry spent fuel storage system
- The MAGNASTOR dry spent fuel storage system Both systems consist of a Transportable Storage Canister (TSC) and concrete Vertical Storage Cask (VSC). The TSC is the CONFINEMENT BOUNDARY for both systems. The TSC is welded and designed to provide confinement of all radionuclides under normal, off-normal, and accident conditions (ref. 1, 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 198 of 260
ATTACHMENT 1 EAL Bases Confinement boundary is defined as the barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC.
CSD-EP-CNS-0101-01 Rev. 000 Page 199 of 260
ATTACHMENT 1 EAL Bases The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification for radiation external to a loaded VSC for a NAC-UMS canister (ref. 1).
The specified ISFSI dose limits are based on surveys taken consistent with the locations specified in the associated Technical Specification (ref. 1, 2).
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 technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. 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.
CNS Basis Reference(s):
- 1. NAC-UMS Certificate of Compliance #1015 Technical Specifications
- 2. MAGNASTOR Certificate of Compliance #1031 Technical Specifications
- 3. NEI 99-01 E-HU1 CSD-EP-CNS-0101-01 Rev. 000 Page 200 of 260
ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (NCS 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 (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.
Containment Barrier thresholds are used as criteria for escalation of the ECL from 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 (Attachment 2). 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 NCS 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 CSD-EP-CNS-0101-01 Rev. 000 Page 201 of 260
ATTACHMENT 1 EAL Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
- The Fuel Clad Barrier and the NCS Barrier are weighted more heavily than the Containment Barrier.
- Unusual Event ICs associated with NCS 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 RG1 has been exceeded.
- The fission product barrier thresholds specified within a scheme reflect plant-specific CNS design and operating characteristics.
- As used in this category, the term NCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of NCS mass to any location- inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the NCS due to the as-designed/expected operation of a relief valve is not considered to be NCS 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 NCS 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 NCS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.
CSD-EP-CNS-0101-01 Rev. 000 Page 202 of 260
ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or NCS EAL:
FA1.1 Alert Any loss or any potential loss of either Fuel Clad or NCS (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS and Containment comprise the fission product barriers. Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and NCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or NCS 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 NCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 CNS Basis Reference(s):
- 1. NEI 99-01 FA1 CSD-EP-CNS-0101-01 Rev. 000 Page 203 of 260
ATTACHMENT 1 EAL Bases Category: 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 Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS and Containment comprise the fission product barriers. Table F-1 (Attachment
- 2) 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 NCS 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 NCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.
CNS Basis Reference(s):
- 1. NEI 99-01 FS1 CSD-EP-CNS-0101-01 Rev. 000 Page 204 of 260
ATTACHMENT 1 EAL Bases Category: 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 third barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS and Containment comprise the fission product barriers. Table F-1 (Attachment
- 2) 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, NCS and Containment barriers
- Loss of Fuel Clad and NCS barriers with potential loss of Containment barrier
- Loss of NCS and Containment barriers with potential loss of Fuel Clad barrier
- Loss of Fuel Clad and Containment barriers with potential loss of NCS barrier CNS Basis Reference(s):
- 1. NEI 99-01 FG1 CSD-EP-CNS-0101-01 Rev. 000 Page 205 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction 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. NCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / NCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Jugement 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 Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category B would be assigned FC Loss B.1, the third Containment barrier Potential Loss in Category D would be assigned CMT P-Loss D.3, etc.
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 CSD-EP-CNS-0101-01 Rev. 000 Page 206 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and NCS 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 NCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,, E.
CSD-EP-CNS-0101-01 Rev. 000 Page 207 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Containment (CMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss
- SG tube RUPTURE
- 1. CSFST Heat Sink-RED Path conditions met
- 2. CSFST Heat Sink-RED Path conditions met
- 1. CSFST Core Cooling-RED Inadequate conditions met None None AND Path conditions met AND Heat AND Restoration procedures not Removal Heat sink is required effective within 15 min. (Note 1)
Heat sink is required C 1. EMF53A/B > Table F-2 column CMT FC Loss 1. EMF53A/B > Table F-2 column 1. EMF53A/B > Table F-2 column None None None Radiation 2. Dose equivalent I-131 coolant NCS Loss CMT Potential Loss
/ NCS activity > 300 µCi/gm Activity
- 1. Containment isolation is required AND EITHER: 1. CSFST Containment-RED Path
- Containment integrity has conditions met D been lost based on 2. Containment hydrogen concentration Emergency Coordinator > 6%
CMT None None None None judgment Integrity 3. Containment pressure > 3 psig
- UNISOLABLE pathway from with < one full train of containment or Bypass Containment to the environment cooling operating per design for exists > 15 min. (Note 1)
- 2. Indications of NCS leakage outside of containment E 1. Any condition in the opinion of the Emergency Coordinator that
- 1. Any condition in the opinion of the Emergency Coordinator that
- 1. Any condition in the opinion of
- 1. Any condition in the opinion of the Emergency Coordinator that
- 1. Any condition in the opinion of the Emergency Coordinator that
- 1. Any condition in the opinion of the Emergency Coordinator that the Emergency Coordinator that EC indicates loss of the fuel clad indicates potential loss of the indicates potential loss of the NCS indicates loss of the containment indicates potential loss of the indicates loss of the NCS barrier Judgment barrier fuel clad barrier barrier barrier containment barrier CSD-EP-CNS-0101-01 Rev. 000 Page 208 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 209 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 210 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
- 1. CSFST Core Cooling-RED Path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
CNS Basis Reference(s):
- 1. EP/1(2)/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling
- 3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling
- 4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A CSD-EP-CNS-0101-01 Rev. 000 Page 211 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
- 1. CSFST Core Cooling-ORANGE Path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
CNS Basis Reference(s):
- 1. EP/1(2)/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling
- 3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling
- 4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A CSD-EP-CNS-0101-01 Rev. 000 Page 212 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
- 2. CSFST Heat Sink-RED Path conditions met AND Heat sink is required Definition(s):
None Basis:
In combination with NCS Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The phrase and heat sink required precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350ºF. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 2)
This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
CSD-EP-CNS-0101-01 Rev. 000 Page 213 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNS Basis Reference(s):
- 1. EP/1(2)/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink
- 3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B CSD-EP-CNS-0101-01 Rev. 000 Page 214 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Loss Threshold:
- 1. EMF53A/B > Table F-2 column FC Loss Table F-2 Containment Radiation - R/hr (EMF53A/B)
Time After S/D CMT Potential FC Loss NCS Loss (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
None Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale).
Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The Table F-2 values, column FC Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown, no sprays and NCS pressure < 1600 psig with ~2% fuel failure (ref. 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 215 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The value is derived as follows:
RP/0/A/5000/015 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown without spray and NCS pressure < 1600 psig x 0.02 (rounded)
(ref. 2).
The radiation monitor reading 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 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.
The radiation monitor reading in this threshold is higher than that specified for NCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the NCS Barrier.
Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
CNS Basis Reference(s):
- 1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration
- 2. RP/0/A/5000/015 Core Damage Assessment
- 3. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A CSD-EP-CNS-0101-01 Rev. 000 Page 216 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Loss Threshold:
- 2. Dose equivalent I-131 coolant activity > 300 µCi/gm Definition(s):
None Basis:
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent I-131 concentration is well above that expected for iodine spikes and corresponds to about 2% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 1).
This threshold indicates that NCS 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.
There is no Potential Loss threshold associated with NCS Activity / Containment Radiation.
CNS Basis Reference(s):
- 1. RP/0/A/5000/015 Core Damage Assessment
- 2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B CSD-EP-CNS-0101-01 Rev. 000 Page 217 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 218 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 219 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 220 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost.
CNS Basis Reference(s):
NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A CSD-EP-CNS-0101-01 Rev. 000 Page 221 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
CNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A CSD-EP-CNS-0101-01 Rev. 000 Page 222 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
- UNISOLABLE NCS leakage
- SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
ECCS (SI) actuation is caused by (ref. 1):
- Pressurizer pressure < 1845 psig
- Containment pressure > 1.2 psig This threshold is based on an UNISOLABLE NCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the NCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE NCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If EOPs direct operators to open the Pressurizer pressure relief valves to implement a core cooling strategy (i.e., a feed and bleed cooldown), then there will exist a reactor coolant flow path from the RCS, past the pressurizer safety and relief valves and into the containment that operators cannot isolate without compromising the effectiveness of the strategy (i.e., for the strategy to be effective, the valves must be kept in the open position); therefore, the flow through the pressure relief line is UNISOLABLE. In this case, the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been eliminated and thus this condition constitutes a loss of the RCS barrier.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.
CSD-EP-CNS-0101-01 Rev. 000 Page 223 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 2. EP/1(2)/A/5000/E-3 Steam Generator Tube Rupture
- 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A CSD-EP-CNS-0101-01 Rev. 000 Page 224 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
- 1. CSFST Integrity-RED path conditions met Definition(s):
None Basis:
The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity - RED path.
CSFST NCS Integrity - Red Path plant conditions and associated PTS Limit Curve A indicates an extreme challenge to the safety function when plant parameters are to the left of the limit curve following excessive NCS cooldown under pressure (ref. 1, 2).
This condition indicates an extreme challenge to the integrity of the NCS pressure boundary due to pressurized thermal shock - a transient that causes rapid NCS cooldown while the NCS is in Mode 3 or higher (i.e., hot and pressurized).
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-P.1 Response to Imminent Pressurized Thermal Shock Condition
- 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B CSD-EP-CNS-0101-01 Rev. 000 Page 225 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 226 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
- 1. CSFST Heat Sink-RED path conditions met AND Heat sink is required Definition(s):
None Basis:
In combination with FC Potential Loss B.2, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The phrase and heat sink required precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350ºF. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 1, 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 227 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the NCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate NCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase NCS pressure to the point where mass will be lost from the system.
CNS Basis Reference(s):
- 1. EP/1(2)/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation/ NCS Activity Degradation Threat: Loss Threshold:
- 1. EMF53A/B > Table F-2 column NCS Loss Table F-2 Containment Radiation - R/hr (EMF53A/B)
Time After S/D CMT Potential FC Loss NCS Loss (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
N/A Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale).
Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The value specified represents, based on core damage assessment procedure RP/0/A/5000/015 Figure 1, the expected containment high range radiation monitor (EMF53A &
B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown with no fuel failure (ref. 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 229 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The value is derived as follows:
RP/0/A/5000/015 Figure 1 Containment Radiation Level vs. Time for RCS Release for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown (rounded) (ref. 2).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the NCS Barrier only.
There is no Potential Loss threshold associated with NCS Activity / Containment Radiation.
CNS Basis Reference(s):
- 1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration
- 2. RP/0/A/5000/015 Core Damage Assessment
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CMT Radiation/ NCS Activity Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 231 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 232 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 233 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the NCS barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is lost.
CNS Basis Reference(s):
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the NCS barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to the inability to reach final safety acceptance criteria before completing all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS 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.
CNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment NCS Potential Loss 6.A CSD-EP-CNS-0101-01 Rev. 000 Page 235 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
- 1. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for NCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., NCS activity values) and IC SU5 for the NCS barrier (i.e., NCS leak rate values).
CSD-EP-CNS-0101-01 Rev. 000 Page 236 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g.,
a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of a standby Site Area Emergency per charging (makeup) pump (NCS Alert per FA1.1 FS1.1 Barrier Potential Loss)
Requires an automatic or manual Site Area Emergency per ECCS (SI) actuation (NCS Barrier Alert per FA1.1 FS1.1 Loss)
There is no Potential Loss threshold associated with NCS or SG Tube Leakage.
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 2. EP/1(2)/A/5000/E-3 Steam Generator Tube Rupture
- 3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A CSD-EP-CNS-0101-01 Rev. 000 Page 237 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 238 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:
- 1. CSFST Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note 1)
Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).
A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,200°F (ref. 1), Fuel Clad barrier is also lost.
This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the NCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered effective if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Coordinator should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
CSD-EP-CNS-0101-01 Rev. 000 Page 239 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
CNS Basis Reference(s):
- 1. EP/1(2)/5000/F-0 Critical Safety Function Status Trees
- 2. EP/1(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling
- 3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling
- 4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A CSD-EP-CNS-0101-01 Rev. 000 Page 240 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Loss Threshold:
None CSD-EP-CNS-0101-01 Rev. 000 Page 241 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Potential Loss Threshold:
- 1. EMF53A/B > Table F-2 column CMT Potential Loss Table F-2 Containment Radiation - R/hr (EMF53A/B)
Time After S/D CMT Potential FC Loss NCS Loss (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
None Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale).
Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The Table F-2 values, column CMT Potential Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown, no sprays and NCS pressure < 1600 psig with ~20% fuel failure (ref. 2).
The value is derived as follows:
RP/0/A/5000/015 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown with no spray and NCS pressure < 1600 psig x 0.20 (rounded)
(ref. 2).
CSD-EP-CNS-0101-01 Rev. 000 Page 242 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and NCS 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 NCS 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 ECL to a General Emergency.
CNS Basis Reference(s):
- 1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration
- 2. RP/0/A/5000/015 Core Damage Assessment
- 3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A CSD-EP-CNS-0101-01 Rev. 000 Page 243 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
- 1. Containment isolation is required AND EITHER:
- Containment integrity has been lost based on EC judgment
- UNISOLABLE pathway from containment to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.
First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of NCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the NCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
CSD-EP-CNS-0101-01 Rev. 000 Page 244 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following the leakage of NCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Following the leakage of NCS 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 R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
CSD-EP-CNS-0101-01 Rev. 000 Page 245 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNS Basis Reference(s):
- 1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A CSD-EP-CNS-0101-01 Rev. 000 Page 246 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
- 2. Indications of NCS leakage outside of containment Definition(s):
None Basis:
ECA-1.2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential NCS leak pathways outside containment include (ref. 1, 2):
- Safety Injection (NI)
- Chemical & Volume Control (NV)
- RCP seals (NC)
- PZR/NCS Loop sample lines (NM)
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the NCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of NCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if NCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.
CSD-EP-CNS-0101-01 Rev. 000 Page 247 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases To ensure proper escalation of the emergency classification, the NCS leakage outside of containment must be related to the mass loss that is causing the NCS Loss and/or Potential Loss threshold A.1 to be met.
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/ECA-1.2 LOCA Outside Containment
- 2. EP/1(2)/A/5000/E-1 Loss of Reactor or Secondary Coolant
- 3. NEI 99-01 CMT Integrity or Bypass Containment Loss CSD-EP-CNS-0101-01 Rev. 000 Page 248 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples 2nd Threshold-Airborne release from Effluent pathway Auxiliary Building Monitor Inside Reactor Vent Building Damper Filt Area Monitor Open valve Open valve Damper 1st Penetration Threshold-Airborne Airborne Monitor Open valve Open valve 1st 2nd Threshold-Threshold- RCS leakage Interface leakage Airborne outside RB release Process Monitor Closed Open valve Open valve Pump Cooling RCP Seal Cooling CSD-EP-CNS-0101-01 Rev. 000 Page 249 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
- 1. CSFST Containment-RED path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 15 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
15 psig is based on the containment design pressure (ref. 2).
If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the NCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
CNS Basis Reference(s):
- 1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees
- 2. UFSAR Section 6.2 Containment Systems
- 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A CSD-EP-CNS-0101-01 Rev. 000 Page 250 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
- 2. Containment hydrogen concentration > 6%
Definition(s):
None Basis:
Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump. (ref. 1).
The Containment Hydrogen Purge and Sample System (VY) is used to monitor the hydrogen concentration inside containment after a severe accident involving core damage. Samples of Containment air are obtained via the containment hydrogen/oxygen sample lines to the Post Accident Containment Sample (PACS) panel located in the auxiliary building. Additionally, the containment hydrogen analyzer system continuously monitors the hydrogen concentration inside containment (ref. 1).
The lower limit of deflagration of hydrogen in air is approximately 6% and is the maximum concentration at which hydrogen igniters can be placed in service (ref. 2).
To generate such levels of combustible gas, loss of the Fuel Clad and NCS barriers must have occurred. With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.
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 potential loss of the Containment Barrier.
CNS Basis Reference(s):
- 1. UFSAR Section 6.2 Containment Systems
- 2. EP/1(2)/A/5000/FR-Z.4 Response to High Containment Hydrogen Concentration
- 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B CSD-EP-CNS-0101-01 Rev. 000 Page 251 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
- 3. Containment pressure > 3 psig with < one full train of containment cooling operating per design for > 15 min. (Notes 1, 10)
Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.
Definition(s):
None Basis:
The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.
One full train of containment cooling operating per design is considered (ref. 1, 2):
- One train of Containment Air Return Fan System (VX-CARF), and
- One train of Containment Spray System (NS)
Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the CARF with a 10-minute delay) within 15 minutes a potential loss of containment exists. At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control. The potential loss of containment applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes.
CSD-EP-CNS-0101-01 Rev. 000 Page 252 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases The potential loss of containment also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the potential loss should be credited regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
CNS Basis Reference(s):
- 2. CNS Technical Specification 3.6.6 Bases
- 4. UFSAR Section 6.2 Containment Systems
- 5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C CSD-EP-CNS-0101-01 Rev. 000 Page 253 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
CNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A CSD-EP-CNS-0101-01 Rev. 000 Page 254 of 260
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
- 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
- Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
CNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A CSD-EP-CNS-0101-01 Rev. 000 Page 255 of 260
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & 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.
The review at CNS was completed using the following Controlling Procedures:
- OP/1(2)/A/6100/003 (Controlling Procedure For Unit Operation)
- OP/1(2)/A/6100/002 (Controlling Procedure For Unit Shutdown)
CSD-EP-CNS-0101-01 Rev. 000 Page 256 of 260
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases CNS Table R-2 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:
If action not performed CNS Procedure Step Action Building/Elevation/Room Mode does this prevent and Step cooldown/ shutdown?
OP/1/A/6100/003 Coordinate with Chemistry Auxiliary Building (Various Enclosure 4.3 Step and Radwaste while Locations) 3.19: U2 Encl. 4.3 performing NC System 1 No Step 3.20 Degas. Radwaste continues Degas OPS thru shutdown & cooldown OP/1/A/6100/003, Align heater vent orifices Turbine Building (Various Enclosure 4.3, Steps per OP/1(2)/B/6250/004 Locations) 3.26, 3.27 and 3.28; (Feedwater Heaters, Vents, 1 No U2 Encl. 4.3, Steps Drains and Bleed Systems) 3.28, 3.29 and 3.30 ,Align VI and SP valves associated with CFPTs OP/1/A/6100/003, Align Auxiliary Steam to Turbine Building (Various Enclosure 4.2, Step CFPTs. Locations) 1 No 3.11: U2 Encl. 4.2 Step 3.10 OP/1&2/A/6100/003, Align "C" Htr Drain Turbine Building (568')
Enclosure 4.2, Steps Pump per 3.13; U2 Encl. 4.2 OP/1(2)/B/6250/004 Step 3.12 (Feedwater Heaters, 1 No Vents, Drains and Bleed Systems) for removal from service OP/1/A/6100/003, Plant activities to ensure Turbine Building (594')
Enclosure 4.2, Step Main Turbine Sealing Steam 1 No 3.14; U2 Encl. 4.2 system responds as Step 3.13 required.
OP/1&2/A/6100/003, Ensure Moisture Separator Turbine Building (619')
Enclosure 4.2, Step Reheater low load valve 3.18: U2 Encl. 4.2 operation per 1 No Step 3.17 OP/1(2)/B/6250/013 (Moisture Separator Reheater Operation)
OP/1/A/6100/003 Secure one Main CFPT per Turbine Building (Mainly Enclosure 4.2 Step OP/1(2)/A/6250/001 594')
3.19: U2 Encl. 4.2 (Condensate and 1 No Step 3.18 Feedwater System)
CSD-EP-CNS-0101-01 Rev. 000 Page 257 of 260
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases If action not performed CNS Procedure Step Action Building/Elevation/Room Mode does this prevent and Step cooldown/ shutdown?
OP/1/A/6100/003, Secure half Main Outside in Main Enclosure 4.2, Step Transformer Cooling Fans Transformer yard 1 No 3.20; U2 Encl. 4.2 and oil pumps Step 3.19 OP/1(2)/A/6100/003 Shutdown the Main Turbine Turbine Building and Enclosure 4.2 Step per OP/1(2)/B/6300/001 Transformer Yard 1 No 3.21: (Turbine generator)
OP/1(2)/A/6100/003, Bypass "F" LP heaters Turbine Building (594" LP Enclosure 4.2, Step Htr Panel). 1 No 3.28 &3.29 OP/1(2)/A/6100/003, Transfer of Aux Steam to Service Bldg. (568')
Enclosure 4.2, Step on line Unit per 3.34 OP/0/B/6250/007 A (Auxiliary Steam System 1 No Alignment) or place Aux Electric Boiler in service per OP/1/B/6250/007 B (Auxiliary Electric Boilers)
OP/1(2)/A/6100/003, Isolate Unit Related Turbine Building. (594')
Enclosure 4.2, Step Steam supply to Aux 1 No 3.37 Steam Header OP/1(2)/A/6100/002, Initiate action to reduce Auxiliary Building (577' Enclosure 4.1, Step VCT pressure per Mechanical Pent. Room) 3.4 OP/1(2)/6200/001 1, 2, 3 No (Chemical and Volume Control System)
OP/1(2)/A/6100/002, Align S/G reverse purge. Both Doghouses Enclosure 4.1, Step 1 No 3.9 OP/1(2)/A/6100/002, Align CM system flow for Turbine Building (619')
Enclosure 4.1, Step Low Pressure cleanup 3 No 3.52 thru Upper Surge Tank.
OP/1(2)/A/6100/002 Shutdown Rod Control Auxiliary Building (594' Enclosure 4.2 or 4.7, System per Electrical Pent Room) 3 No Step 3.3 OP/1(2)/6150/008 (Rod Control)
OP/1(2)/A/6100/002 Chemistry obtains Auxiliary Building (543' Enclosure 4.2 or 4.7 samples to ensure boron Sample Lab)
Step 3.10 concentration good to 3 No allow NCS cooldown to begin CSD-EP-CNS-0101-01 Rev. 000 Page 258 of 260
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases If action not performed CNS Procedure Step Action Building/Elevation/Room Mode does this prevent and Step cooldown/ shutdown?
OP/1(2)/A/6100/002 Perform PZR PORVs Auxiliary. Building (577',
Enclosure 4.2 or 4.7, stroke testing per IF performing IWVR Step 3.22 PT/1(2)/A/4200/023 A Containment 635' as 3 No (NC valve Inservice Test) well)
OP/1(2)/A/6100/002 Support placing N2 Auxiliary Building Enclosure 4.2 or 4.7 Cover gas on NCDT per (Various Locations on Step 3.23 OP/1(2)/A/6500/014 577' & 560') 3 No (Operations Controlled Liquid Waste Systems)
OP/1(2)/A/6100/002, Removing CLAs from Auxiliary Building (577' &
Enclosure 4.2 or 4.7 service per 560' Ess. MCC Bkrs)
Step 3.31 OP/1(2)/A/6200/009 3 No (Cold Leg Accumulator Operation).
OP/1&2/A/6100/002 Remove CAPT and one Auxiliary Building. (577' &
Enclosure 4.2 or 4.7, Motor Driven CA Pump 560 Ess MCC Bkrs)
Step 3.45.1 from service per 4 No OP/1(2)/A/6250/002 (Auxiliary Feedwater System).
OP/1(2)/A/6100/002 Open NCS Loop Suction Auxiliary. Building. (577' Enclosure 4.2 or 4.7, Vlvs for train of ND to be & 560' Ess. MCC Bkrs)
Step 3.46.3 placed in service per 4 Yes OP/1(2)/A/6200/004 (Residual Heat removal System).
OP/1(2)/A/6100/002 Rack out appropriate NI and Auxiliary Building (577' &
Enclosure 4.2 or 4.7, NV Pump Motor Bkrs per 560 Electrical Pent Step 3.48.2 OP/0/A/6350/010 (Operation Rooms) 4 Yes of Station Breakers and Disconnects)
OP/1(2)/A/6100/002 Support placing first train of Auxiliary Building (577' or Enclosure 4.2 or 4.7 ND in service per 560' Ess MCC Bkr s)
Step 3.52.2 OP/1(2)/A/6200/004 4 Yes (Residual Heat removal System)
CSD-EP-CNS-0101-01 Rev. 000 Page 259 of 260
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2/H-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Mode Rm 478 (1EMXA) Rm 469 (2EMXA) 4 Rm 496 (1ETA) Rm 486 (2ETA) 4 Auxiliary 577 Rm 496 (1EMXS) Rm 486 (2EMXS) 4 AB-577', JJ-57 (1MXK) AB-577', JJ-57 (2MXK) 4 Rm 330 (1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1ETB) Rm 362 (2ETB) 4 Rm 372 (1EMXD) Rm 362 (2EMXD) 4 Plant Operating Procedures Reviewed
- 1. OP/1(2)/A/6100/003 (Controlling Procedure for Unit Operation)
- 2. OP/1(2)/A/6100/002 (Controlling Procedure for Unit Shutdown)
- 3. OP/1(2)/B/6250/004 (Feedwater Heaters, Vents, Drains and Bleed Systems)
- 4. OP/1(2)/B/6250/013 (Moisture Separator Reheater Operation)
- 5. OP/1(2)/A/6250/001 (Condensate and Feedwater System)
- 6. OP/1(2)/B/6300/001 (Turbine generator)
- 7. OP/0/B/6250/007 A (Auxiliary Steam System Alignment)
- 8. OP/1/B/6250/007 B (Auxiliary Electric Boilers)
- 9. OP/1(2)/6200/001 (Chemical and Volume Control System)
- 10. OP/1(2)/6150/008 (Rod Control)
- 11. PT/1(2)/A/4200/023 A (NC valve Inservice Test)
- 12. OP/1(2)/A/6500/014 (Operations Controlled Liquid Waste Systems)
- 13. OP/1(2)/A/6200/009 (Cold Leg Accumulator Operation)
- 14. OP/1(2)/A/6250/002 (Auxiliary Feedwater System)
- 15. OP/1(2)/A/6200/004 (Residual Heat removal System)
- 16. OP/0/A/6350/010 (Operation of Station Breakers and Disconnects)
CSD-EP-CNS-0101-01 Rev. 000 Page 260 of 260
INTENTIONAL BLANK PAGE Rev. 150 E-1 December 2019
Catawba Nuclear Station Emergency Plan Section E - Notification Methodology E. Notification Methodology E.1 Notification of Response Organization This section identifies specific emergency responses and related criteria that specify when these measures are to be implemented. Emergency measures discussed in this section include notification of and activation of the emergency organization; assessment actions; corrective and protective actions.
E.2 Activation of Emergency Organization This section describes the necessary communication steps to be taken to alert or activate emergency personnel for each class of emergency listed in Section D. The notification format and message authentication technique to off-site authorities shall be in accordance with AD-EP-ALL-0304, State and County Notifications.
E.2.a. Notification of Unusual Event The Shift Manager on duty is to be notified immediately of all initiating conditions indicative of an "Unusual Event" in process or that has occurred which indicates a potential degradation in the level of safety of the plant. (See Section D for examples of initiating conditions in this classification.)
NOTE: This emergency classification is further defined in AD-EP-ALL-0101, Emergency Classification.
The Shift Manager assumes the functions of the Emergency Coordinator and shall ensure that all actions required by any initiating Emergency Procedure have been performed and that all actions necessary for the protection of persons and property are being taken.
The Shift Manager shall assure notification of:
- 1. Station Manager
- 2. Site Vice President
- 3. Chief Nuclear Officer for any initiating condition in this classification listed in Section D.
Rev. 150 E-1 December 2019
The Shift Manager shall assure prompt notification of Federal, State and Local off-site authorities:
- 1. North Carolina Warning Point (Raleigh, NC)
- 2. South Carolina Warning Point (Columbia, SC)
- 3. York County Warning Point (Rock Hill, SC)
- 4. Gaston County Warning Point (Gastonia, NC)
- 5. Mecklenburg County Warning Point (Charlotte, NC)
- 6. NRC Operations Center (Rockville, MD)
Notification format and message authentication technique to off-site authorities shall be in accordance with applicable Catawba Nuclear Station Emergency Response Procedures.
The Shift Manager shall augment on-shift resources to assess and respond to the emergency situation as needed to ensure the protection of persons and property.
The Shift Manager will assess the emergency condition and determine the need to remain in a Notification of Unusual Event, escalate to a more severe class or close out the emergency.
The Manager, Nuclear Support Services, will close out the Emergency with verbal summary to off-site authorities, notified above, followed by an LER or written summary within 30 days.
The actions required for this emergency class are performed by station personnel.
Outside organizations (NRC, state and local officials) are notified of the event for information. Unless deemed necessary by the Emergency Coordinator, the Emergency Response Organization is not activated for this emergency class.
If an Unusual Event occurs, a station representative calls the NRC, the State, appropriate local officials, Corporate Communications, and others as applicable. The Corporate Communications representative notifies media representatives and public officials per established public information procedures.
E.2.b Alert The Shift Manager on duty is to be notified immediately of all initiating conditions indicative of an "Alert" classification in process or that 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. (See Section D for examples of initiating conditions in this classification.)
I NOTE: This Emergency Classification is further defined in AD-EP-ALL-0101, Emergency Classification.
Rev. 150 E-2 December 2019
The Shift Manager shall ensure that all actions required by any initiating Emergency Procedure have been performed and that all actions necessary for the protection of persons and property are being taken.
NOTE: The Shift Manager assumes the function of the Emergency Coordinator until the arrival of the Station Manager or designee at which time the Station Manager or designee assumes the responsibility of the Emergency Coordinator.
The Shift Manager shall assure notification and activation of the Emergency Response Organization for any initiating condition in this classification listed in Section D.
The Emergency Response Organization personnel will be notified by text and/or voice message upon the initial emergency declaration via a mass communication system using AD-EP-ALL-0301, Activation of the Emergency Response Organization Notification System (ERONS). Redundant notification is provided by the on-site public address system, Nuclear Callout System and/or an automated telephone system which will allow timely alerting of Emergency Response Organization personnel.
The Emergency Coordinator shall assure prompt notification of Federal, State and Local off-site authorities:
- 1. North Carolina Warning Point (Raleigh, NC)
- 2. South Carolina Warning Point (Columbia, SC)
- 3. York County Warning Point (Rock Hill, SC)
- 4. Gaston County Warning Point (Gastonia, NC)
- 5. Mecklenburg County Warning Point (Charlotte, NC)
- 6. NRC Operations Center (Rockville, MD)
Notification format and message authentication technique to off-site authorities shall be in accordance with applicable Catawba Nuclear Station Emergency Response Procedures.
The Emergency Coordinator shall augment on-site resources by notification and activation of the Emergency Response Organization in accordance with AD-EP-ALL-0301, Activation of the Emergency Response Organization Notification System (ERONS).
The Emergency Coordinator in the Technical Support Center will assess and respond to the emergency by:
- 1. Dispatching on-site monitoring teams with associated communication equipment in accordance with Catawba Nuclear Station Radiation Protection procedures.
- 2. Providing periodic plant status updates to off-site authorities (at least every hour or as agreed otherwise).
Rev. 150 E-3 December 2019
- 3. Providing periodic meteorological assessments to off-site authorities and, if any releases are occurring, dose estimates for actual releases.
NOTE: These functions will be provided through the EOF when operational.
The Emergency Coordinator will assess the emergency condition and determine the need to remain in an Alert status, escalate to a more severe class, reduce the emergency class or close out the emergency.
The EOF Director or designee, will close out the emergency with a verbal summary to off-site authorities followed by an LER or written summary within 30 days.
E.2.c. Site Area Emergency The Shift Manager on duty is to be notified immediately of all initiating conditions indicative of a "Site Area Emergency" in process or which 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. (See Section D for examples of initiating conditions in this classification.)
NOTE: This Emergency Classification is further defined in AD-EP-ALL-0101, Emergency Classification.
The Shift Manager shall ensure that all actions required by the initiating Emergency Procedure have been performed and that all actions necessary for the protection of persons and property are being taken.
NOTE: The Shift Manager assumes the function of the Emergency Coordinator until the arrival of the Station Manager or designee at which time the Station Manager or designee assumes the responsibility of the Emergency Coordinator.
The Shift Manager shall assure notification and activation of the Emergency Response Organization for any initiating condition in this classification listed in Section D.
The Emergency Coordinator shall assure prompt notification of Federal, State and Local off-site authorities:
- 1. North Carolina Warning Point (Raleigh, NC)
- 2. South Carolina Warning Point (Columbia, SC)
- 3. York County Warning Point (Rock Hill, SC)
- 4. Gaston County Warning Point (Gastonia, NC)
- 5. Mecklenburg County Warning Point (Charlotte, NC)
- 6. NRC Operations Center (Rockville, MD)
Rev. 150 E-4 December 2019
Notification format and message authentication technique to off-site authorities shall be in accordance with applicable Catawba Nuclear Station Emergency Response Procedures.
I The Emergency Coordinator shall augment on-site resources by notification and activation of the Emergency Response Organization in accordance with AD-EP-ALL-0301, Activation of the Emergency Response Organization Notification System (ERONS).
The Emergency Response Organization personnel will be notified by text and/or voice message upon the initial emergency declaration via a mass communication system using AD-EP-ALL-0301 (Activation of the Emergency Response Organization Notification System (ERONS). Redundant notification is provided by the on-site public address system, Nuclear Callout System and/or an automated telephone system which will allow timely alerting of Emergency Response Organization personnel.
The Emergency Coordinator may order the evacuation of non-essential station personnel to an Evacuation-Relocation Site if the emergency situation warrants.
The Emergency Coordinator in the Technical Support Center will assess and respond to the emergency by:
- 1. Dispatching the On-site and Off-site Monitoring Teams with associated communications.
- 2. Providing meteorological and dose estimate information to off-site authorities for actual releases via a dedicated individual or automated data transmission.
- 3. Providing release and dose projections based on available plant condition information and foreseeable contingencies to off-site authorities.
- 4. Providing a dedicated individual for plant status updates to off-site authorities.
- 5. Providing technical staff on-site available for consultation with the NRC and State on a periodic basis.
NOTE: These functions will be provided through the EOF when it is operational.
The Emergency Coordinator, in coordination with the EOF Director, will assess the emergency condition and determine the need to remain in Site Area Emergency, escalate to a more severe class, reduce the emergency class or close out the emergency.
The EOF Director will close out or recommend reduction of the emergency class by briefing the off-site authorities at the EOF or by phone if necessary, followed by an LER or written summary within thirty days.
Rev. 150 E-5 December 2019
E.2.d General Emergency The Shift Manager on duty is to be notified immediately of all initiating conditions indicative of a "General Emergency" in process or which 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. (See Section D for examples of initiating conditions in this classification.)
NOTE: This Emergency Classification is further defined in AD-EP-ALL-0101, Emergency Classification.
The Shift Manager shall ensure that all actions required by the initiating Emergency Procedure have been performed and that all actions necessary for the protection of persons and property are being taken.
NOTE: The Shift Manager assumes the function of the Emergency Coordinator until the arrival of the Station Manager or designee at which time the Station Manager or designee assumes the responsibility of the Emergency Coordinator.
The Shift Manager shall assure notification and activation of the Emergency Response Organization for any initiating condition in this classification listed in Section D.
The Emergency Coordinator shall assure prompt notification of Federal, State and Local off-site authorities:
- 1. North Carolina Warning Point (Raleigh, NC)
- 2. South Carolina Warning Point (Columbia, SC)
- 3. York County Warning Point (Rock Hill, SC)
- 4. Gaston County Warning Point (Gastonia, NC)
- 5. Mecklenburg County Warning Point (Charlotte, NC)
- 6. NRC Operations Center (Rockville, MD)
Notification format and message authentication technique to off-site authorities shall be in accordance with applicable Catawba Nuclear Station Emergency Response Procedures.
The Emergency Coordinator shall augment on-site resources by notification and activation of the Emergency Response Organization in accordance with AD-EP-ALL-0301, Activation of the Emergency Response Organization Notification System (ERONS).
Rev. 150 E-6 December 2019
The Emergency Response Organization personnel will be notified by text and/or voice message upon the initial emergency declaration via a mass communication system using AD-EP-ALL-0301, Activation of the Emergency Response Organization Notification System (ERONS). Redundant notification is provided by the on-site public address system, Nuclear Callout System and/or an automated telephone system which will allow timely alerting of Emergency Response Organization personnel.
The Emergency Coordinator shall order the evacuation of all non-essential station personnel to an Evacuation-Relocation Site.
The Emergency Coordinator, in the Technical Support Center, will assess and respond to the emergency by:
- 1. Dispatching the On-Site and Off-Site Monitoring Teams with associated communications.
- 2. Providing meteorological and dose estimate information to off-site authorities for actual releases via a dedicated individual or automated data transmission.
- 3. Providing release and dose projections based on available plant condition information and foreseeable contingencies to off-site authorities.
- 4. Providing a dedicated individual for plant status updates to off-site authorities and periodic press briefings.
- 5. Providing technical staff on-site available for consultation with the NRC and State on a periodic basis.
NOTE: These functions will be provided through the EOF when it is operational.
The Emergency Coordinator shall make a recommendation to the off-site authorities to evacuate and/or shelter affected zones in accordance with AD-EP-ALL-0109, Offsite Protective Actions Recommendations.
The Emergency Coordinator, in coordination with the EOF Director and Off-Site Agencies, will assess the emergency condition and determine the need to remain in a General Emergency or terminate the emergency and enter into Recovery.
Rev. 150 E-7 December 2019
The EOF Director will terminate the emergency class and recommend entry into recovery by briefing the off-site authorities at the Emergency Operations Facility, or by phone if necessary, followed by an LER or written summary within thirty days.
E.3 Emergency Message Format (Initial)
Figure E-1, Emergency Notification contains information about the class of emergency, whether a release is taking place, the potentially affected areas and whether protective actions may be necessary.
E.4 Emergency Message Format (Follow-Up)
Figure E-1, Emergency Notification contains provisions for follow-up information if it is known and appropriate.
E.5 State and Local Organizations-Disseminating Public Information State and Local plans provide for disseminating information in Initial and Follow-up Messages to the public. (See state and local plans).
E.6 Alert and Notification System The Alert and Notification System for Catawba Nuclear Station will include an acoustic alerting signal, tone alert radios for special facilities and notification of the public by the Emergency Alert System (EAS). The system is designed to meet the acceptance criteria of Section B of Appendix 3, NUREG-0654, FEMA-REP-1, Rev. 1. As a back-up, State and Local plans maintain the alert mechanism via emergency vehicles, automated dialing systems, PA Systems, etc. to also alert the public to monitor commercial broadcasts for emergency information.
See Appendix 3, Alert and Notification System Plan.
Each county will control the activation of the sirens within its boundaries.
Duke Energy will cooperate with FEMA and the state/local governments in their sampling of the residents to assess the ability to hear the alerting signal, the public's awareness of the meaning of the prompt notification message, and the availability of emergency information.
The siren system will be tested and maintained in accordance with the schedule as specified in Appendix 3.
The EAS System is the primary notification system. Backups include the use of county vehicles with audio equipment, county automated dialing systems, and other media communications.
Rev. 150 E-8 December 2019
E.7 Supporting Information for Public Information Messages The portion of Figure E-1 in which protective action recommendations are made assists the state and local authorities in preparing messages for the public's information via the EAS (Emergency Alert System).
EAS message formats are described in the North Carolina and South Carolina Emergency Plans.
Rev. 150 E-9 December 2019
Figure E-1 Page 1 of 1 NUCLEAR POWER PLANT EMERGENCY NOTIFICATION FORM MESSAGE# _ __ Confirmation Phone#: _ _ _ _ _ _ _ _ AUTHENTICATION CODE#: _ _ _ __
Lines 1 - 6 are required for INITIAL Notifications
- 1. EVENT: D DRILL D ACTUAL DECLARATION D TERMINATION (ONLY Lines 1, 2, & 4 required)
- 2. AFFECTED SITE:
CATAWBA
- 3. EMERGENCY CLASSIFICATION D UNUSUALEVENT D ALERT D SITE AREA EMERGENCY D GENERALEMERGENCY
- 4. EAL#_ _ _ _ __ Declaration Date: _ _/_ _/_ _ Time: _ __
Termination Date: _ _/_ _/_ _ Time: _ _ _ (mark "N/A" for EAL# & Description)
EAL DESCRIPTION: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
- 5. RELEASE TO THE ENVIRONMENT (caused by the emergency): D NONE D IS OCCURRING D HAS OCCURRED
- 6. PROTECTIVE ACTION RECOMMENDATIONS:
D NONE D EVACUATE: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
D SHELTER: - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
D CONSIDER THE USE OF Kl (POTASSIUM IODIDE) IN ACCORDANCE WITH ORO PLANS AND POLICIES D OTHER:
Lines 7-11 are NOT required for INITIAL notifications. Lines 7-11 may be provided separately for follow-up notifications.
- 7. PROGNOSIS: Upgrade in classification or PAR change is likely before the next follow-up notification D Yes D No
- 8. SITE UNIT(S) STATUS:
AFFECTED UNIT DYES Unit 1 - _ _ _ _ % Power Shutdown: Date _ /_ _/_ _Time _ _ __
DYES Unit 2 - % Power Shutdown: Date _ /_ _/_ _Time _ _ __
- 9. METEOROLOGICAL DATA:
Wind direction from: _ _ _ degrees Wind Speed: _ _ _ mph Precipitation: _ _ _ inches Stability Class: D A D B DC DD DE DF DG Lines 10 - 11 are completed for follow-up notifications, IF Line 5 IS OCCURRING or HAS OCCURRED is selected
- 10. AIRBORNE RELEASE CHARACTERIZATION: D GROUND D MIXED D ELEVATED MAGNITUDE UNITS: D Ci D Ci/sec D µCi/sec Noble Gases: _ _ _ __ Iodines: _ _ _ __ Particulates: _ _ _ __
- 11. DOSE PROJECTION: Projection period: _ _ _ _ Hours Estimated Release Duration _ _ _ Hours Performed: DISTANCE TEDE (mrem) Thyroid CDE (mrem)
Date _ / __/__ Site Boundary Time:
2 Miles 5 Miles 10 Miles
- 12. REMARKS (As Applicable): _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
- 13. APPROVED BY: _ _ _ _ _ _ _ _ _ _ _TITLE: _ _ _ _ _ _ _ _ Date _ /_ _/__Time _ _ __
- 14. NOTIFIED BY: Date _ /_ _/_ _Time _ _ __
- 15. RECEIVED BY (ORO use only): _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Date _ /_ _/__ Time _ _ __
EM-78 / Nuclear Power Facility Emergency Notification Form/ March 2016 revision Rev. 150 E-10 December 2019
INTENTIONAL BLANK PAGE Catawba Nuclear Station Emergency Plan Section I - Accident Assessment I. ACCIDENT ASSESSMENT To assure the adequacy of methods, systems and equipment for assessing and monitoring actual or potential off-site consequences of a radiological emergency condition.
I.1 Emergency Action Level Procedures Emergency Action Level procedures have been established in accordance with NUMARC/NESP-007 (Rev. 2) that was approved by the NRC in Revision 3 of Regulatory Guide 1.101, and subsequent guidance provided in NRC Bulletin 2005-02, the guidance endorsed in RIS 2006-12 and to support implementation of NEI 03-12. 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).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012, CNS conducted an EAL implementation upgrade project that produced the EALs discussed Section D. See Discussion in Section D.
AD-EP-ALL-0101, Emergency Classification and CSD-EP-CNS-0101-02, EAL Wallcharts will identify the system parameter and effluent parameter values which can be used to determine the emergency condition.
I.2 On-site Capability and Resources to Provide Initial Values and Continuing Assessment I.2.a. Post Accident Sampling Changes have been made to reflect the NRC approved License Amendment Request received from the NRC via letter dated 9/11/01. The NRC issued Amendments No. 193 (Facility Operating License NPF-35) and No. 185 (Facility Operating License NPF-52). The amendments delete TS section 5.5.4, "Post Accident Sampling," for Catawba Nuclear Station, Units 1 and 2, and thereby eliminate the requirements to have and maintain the post-accident sampling systems (PASS - Palss/Pacs).
Chemistry Management Procedure 3.4.12 describes current post accident contingency plans for obtaining NC and ND/Containment Sump samples. It indicates that procedures OP/1(2)/A/6200/011 have been revised to take required samples during accident conditions utilizing the NM Sampling System. The samples are cooled by the normal NM sample HXs cooled by YN, thus eliminating the need for the PALS Sample Cooler.
The following procedures are in place to assess core damage and take containment atmosphere samples under accident condition. Emergency Planning Implementing Procedure RP/0/A/5000/015 is used to assess core damage using EMF 53 response. HP/0/B/1001/018 was revised to allow for use of EMF 38 and 39 containment atmosphere sampling capability under emergency conditions.
Rev. 148 I-1 December 2019
Also as a result of NRC License Amendments 193/185, OP/0/B/6200/021, "PALSS Operation for Accident Sampling," has been deleted from the Emergency Plan as an Emergency Plan Implementing Procedure. Procedures OP/1(2)/A/6200/011 are not EPIPs or a part of the Emergency Plan. They are listed in this section for reference purposes only. Also, Emergency Plan Implementing Procedures HP/1/B/1009/017, "Unit 1 Nuclear Post-Accident Containment Air Sampling System Operating Procedure," and HP/2/B/1009/017, "Unit 2 Nuclear Post-Accident Containment Air Sampling System Operating Procedure," have been deleted. HP/0/B/1001/018, "RP Compliance Sampling," is not an EPIP or a part of the Emergency Plan. It is listed in this section for reference purposes only. (PIPs C-01-00384, C-01-04478)
I.2.b. Radiation and Effluent Monitors Radiological monitoring capabilities include process and effluent monitoring systems (UFSAR 11.5); area monitoring system (UFSAR 12.3.4); plus station portable monitoring instruments, laboratory counters and analyzers (UFSAR 12.5.2), including emergency high-range instruments with a range up to 1000 R/hr and air samplers.
In addition, there are two (2) high range containment monitors, one (1) high range unit vent monitor, and four (4) steam line monitors per unit.
I.2.c In-plant Iodine Instrumentation Silver Zeolite radioiodine or equivalent sampling cartridges are used for sampling air when the presence of noble gases is suspected. Radiation Protection personnel are knowledgeable in the appropriate station procedures required and are trained in the equipment required to determine airborne iodine concentrations in the plant under all conditions. Procedures to determine airborne iodine concentrations will cover analyses to be done if counting room capabilities are not available.
I.3.a/ Method For Determining Release Source Term I.3.b Procedures AD-EP-ALL-0203, AD-EP-CNS-0203, HP/0/B/1009/006, HP/0/B/1009/007, HP/0/B/1009/014, and AD-EP-ALL-0202 are used on-shift, in the TSC and/or EOF for the calculation of potential off-site doses based on a Design Basis Accident, release of primary coolant, or release of GAP activity situation scaled to actual containment monitor readings.
Provisions for use of actual source terms exist in the procedures.
The magnitude of the release is based on actual effluent monitoring readings, plant system parameters (containment pressure), area meteorology and the duration of the release.
I.4 Effluent Monitor Readings Vs On-site/Off-site Exposure The procedures referenced in I.3.a/I.3.b establish the relationship between effluent monitor readings and on-site/off-site exposures and contamination for various meteorological conditions.
Rev. 148 I-2 December 2019
I.5 Meteorological Information Availability Meteorological information will be available to the Emergency Operations Facility, the Technical Support Center, the Control Room through use of the Station Operator Aid Computer (OAC) and by direct telephone communication. Meteorological information will be available to the NRC through the Emergency Response Data System (ERDS), the FTS Health Physics Network (HPN) phone or by direct telephone communications with the individual responsible for making off-site dose assessments either at the Technical Support Center or the Emergency Operations Facility.
Meteorological information will also be given to both the county Emergency Operations Centers, the State of South Carolina and the State of North Carolina during initial and follow-up information via the message format in Figure E-1.
I.6 Release Rates/Projected Dose For Off-scale Instrumentation If instrumentation used for dose assessment are off-scale or inoperable, dose rates within the Reactor Building will be determined using procedure HP/0/B/1009/006, Alternative Method for Determining Dose Rate within the Reactor Building.
I.7/ Field Monitoring Within E.P.Z.
I.8 Field monitoring within the Catawba Emergency Planning Zone will be performed in accordance with AD-EP-ALL-0202, Field Monitoring During Declared Emergency and AD-EP-CNS-0203, CNS Site Specific Field Monitoring Information.
Four off-site field monitoring teams are comprised from station personnel and are under the direction of the Field Monitoring Coordinator. Procedure AD-EP-ALL-0203, Field Monitoring During Declared Emergency and AD-EP-CNS-0203, CNS Site Specific Field Monitoring, describes how to obtain the vehicles to be used, routes to be used, sampling and monitoring equipment to be used, locations of TLD's and directions for taking KI tablets.
In addition, one on-site (out of plant) survey team is available from the Operations Support Center.
An emergency radio system is available for the field monitoring teams to use to relay information to the Control Room/TSC/EOF. The states will be able to monitor the results of the field monitoring teams.
I.9 Detect and Measure Radioiodine Concentration in the EPZ Appropriate instrumentation to measure radioactivity in counts per minute (cpm) and determine dose rate in mrem/hr shall be used for detection and measurement of radioiodine concentration.
The air sample will be taken with a Portable Air Sampler equipped with a Silver Zeolite or equivalent cartridge and particulate filter. Air sampling results will be obtained through the use of a portable single channel Analyzer and appropriate gamma sensitive detector OR a count rate meter utilizing direct corrected count rate (ccpm) of Silver Zeolite or equivalent cartridge cross referenced against an estimated Iodine 131 µCi/cc (microcuries per cubic centimeter) concentration attachment.
Rev. 148 I-3 December 2019
Interference from the presence of noble gas and background radiation shall not decrease the minimum detectable activity of 1E-7 µCi/cc (microcuries per cubic centimeter) under field conditions.
These samples taken by the off-site monitoring teams will be evaluated further by one of the available laboratory facilities described in Section C.3. A multi-channel analyzer will be used to perform this evaluation.
I.10 Relationship Between Contamination Levels and Integrated Dose/Dose Rates Provisions for assessing contamination levels, water, and air to dose rates for key isotopes is found in the Offsite Dose Calculation Manual (ODCM).
I.11 Plume Tracking The states of North Carolina and South Carolina have arrangements to locate and track an airborne plume of radioactive materials. Duke Energy will have monitoring teams in the field, fixed TLD sites and the capability for obtaining airborne monitoring to assist in plume tracking Rev. 148 I-4 December 2019
INTENTIONAL BLANK PAGE Catawba Nuclear Station Emergency Plan Section J - Protective Response J. PROTECTIVE RESPONSE To assure that a range of protective actions is available for the plume exposure pathway for emergency workers and the public. Guidelines for protective actions during an emergency, consistent with Federal guidance, are developed and in place and protective actions for the ingestion exposure pathway appropriate to the locale have been developed.
To protect onsite personnel during hostile action and ensure the continued ability to safely shutdown the reactor and perform the functions of the emergency plan a range of protective actions are in place.
J.1. On-site Alerting and Notification a-d The means and time required to warn, alert and/or notify employees not having emergency assignments (non-essential), visitors, contractor and construction personnel and other individuals who may be on or passing through the owner-controlled area are described in Emergency Response Procedure RP/0/A/5000/010, Conducting a Site Assembly/Evacuation.
Methods to notify and alert onsite personnel (essential and non-essential) during hostile action activities are described in RP/0/B/5000/026, Site Response to Security Events, AP/0/A/5500/046, Hostile Aircraft Activity and RP/0/A/5000/010, Conducting a Site Assembly or Preparing the Site for an Evacuation.
J.2 Evacuation Routes and Transportation The Operations Shift Manager/Emergency Coordinator or designee uses station and local area maps, information available from meteorological tower instrument readouts and current radiological data for determining the evacuation route. Provisions for evacuation of on-site individuals include evacuation by private automobile. Personnel would then drive along Concord Road (SR 1132) west (which is not in one of the prevailing wind directions) to SC Highway 274. Personnel would then drive either South approximately 11 miles and assemble at the Duke Energy, York Operations Center or North approximately 10 miles to the Duke Energy, Allen Plant. The relocation site will have decontamination and contamination control capability and equipment in the event it is needed.
Evacuation by automobile requires 15 to 30 minutes depending on which Relocation Site is chosen.
High traffic density is not considered in estimating evacuation times due to the relatively untraveled area selected for the site (UFSAR Section 2.2.2.1).
J.3 Personnel Monitoring Radiation Protection emergency personnel survey teams equipped with portable monitoring instruments will monitor employees, visitors, construction workers and vehicles for contamination at the Relocation Sites. Monitoring will be performed in accordance with procedure HP/0/B/1009/005 Personnel/Vehicle Monitoring for Emergency Conditions.
Rev 148 J-1 December 2019
J.4 Site Evacuation Procedures - Decontamination/Non-Essential Personnel Criteria Non-essential personnel may be evacuated from the plant site in the event of a Site Area Emergency and will be evacuated in the event of a General Emergency. Provisions are made for the decontamination of vehicles and personnel at an off-site location if the situation should warrant.
All members of the general public who are on-site must be evacuated if there is a possibility they may exceed either of the following limits:
External Radiation Level = 2 mrems/hr Airborne Radioactivity = 1 times DAC for an unrestricted area During hostile threat conditions expedited relocation of personnel to locations away from the hazards area are performed in accordance with RP/0/B/5000/026, Site Response to Security Events, AP/0/A/5500/046, Hostile Aircraft Activity and RP/0/A/5000/010, Conducting a Site Assembly or Preparing the Site for an Evacuation.
J.5 Personnel Accountability Within thirty minutes of a Site Assembly, all persons within the Protected Area of Catawba Nuclear Station can be accounted for and any person(s) determined to be missing, will be identified by name.
RP/0/A/5000/010 provides for the accounting of personnel (on site) continuously thereafter.
During hostile threat conditions personnel accountability is performed in accordance with RP/0/A/5000/010, Conducting a Site Assembly or Preparing the Site for an Evacuation.
J.6 Protective Measures - Breathing Apparatus, Protective Clothing, KI Protective equipment and supplies will be distributed (as needed) to personnel remaining or arriving on site during the emergency to minimize the effects of radiological exposures or contamination.
Protective measures to be utilized are as follows:
- Protective measures will be utilized to minimize the ingestion and/or inhalation of radionuclides and to maintain internal exposure below the limits specified in 10CFR20, Appendix B.
- Engineering (ventilation) controls are utilized in the TSC and Control Room to control concentrations of radioactive material in air. Otherwise, when not practical to apply process or other engineering controls to limit intakes of radioactive material in air, one or more of the following protective measures will be utilized:
- Control of access
- Limitation of exposure times
- Use of individual respiratory protection equipment. Specific positions within the TSC and OSC are required to be respirator qualified. These positions are:
TSC - Operations Manager, Assistant Operations Manager, Engineering Manager, Mechanical Engineer, Electrical Engineer and Reactor Engineer OSC - All positions except the OSC Log Keeper Rev 148 J-2 December 2019
- Self-contained breathing apparatus will be used in areas that are deficient in oxygen or when fighting fires. Respiratory protective equipment will be issued by Radiation Protection or Safety and Health Services. SCBA's are available with other firefighting equipment for use by the station fire brigade.
- Individual Thyroid Protection - Protective measures will be utilized to minimize the ingestion and/or inhalation of radioactive iodine. However, if an unplanned incident involves the accidental or potential ingestion or inhalation of radioactive iodine, Potassium Iodide Tablets (KI) are available for distribution by AD-EP-ALL-0204 (Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release).
- Use of Protective Clothing - Protective clothing will be issued when contamination levels exceed 1000 dpm/100 cm² beta-gamma and 20 dpm/100 cm² alpha of smearable contamination. Protective clothing items are located in the Change Rooms inside the Radiation Control Area, available for emergency use. Special fire-fighting protective clothing and equipment is available in designated station supply storage areas for use by fire brigade personnel.
J.7 Protective Action Recommendations The Emergency Coordinator (Operations Shift Manager or Station Manager) or the EOF Director shall be responsible for contacting the state and/or local governments to give prompt notification for implementing protective measures within the plume exposure pathway.
Protective Action Guides are adopted from EPA 400-R-92-001 and are shown in Figure J-2. A flowchart to aid the Emergency Coordinator in making Protective Action Recommendations is also shown in AD-EP-ALL-0109, Off Site Protective Actions Recommendations.
As described in section B.4, the Emergency Coordinator and the EOF Director are responsible for making protective action recommendations. Prior to activation/operation of the EOF, the Emergency Coordinator will be responsible for making these recommendations. After activation of the EOF, the EOF Director assumes this responsibility. Protective action recommendations will be provided to the off-site authorities (states and counties) who are responsible for implementing public protective actions. Refer to AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment, for protective action recommendations concerning the use of KI by the public. The pre-established warning message format (Figure E-1) will be used in transmitting the recommendations.
Rev 148 J-3 December 2019
The mechanism for making dose projections upon EOF activation is as follows:
The Radiological Assessment Manager is responsible for making dose projections on a periodic basis. These calculations will use existing plant procedures to calculate projected dose to the population-at-risk for either potential or actual release conditions. For conditions in which a release has not occurred but fuel damage has taken place and radiation levels in the containment building atmosphere are significant, a scoping analysis will be performed to determine what recommendations would be made if containment integrity were lost at that time. The analysis will be based upon a design leak rate and upon a projected penetration failure indicated by a hole size of certain diameter. This analysis will include the use of actual containment pressure, realistic meteorology, and actual source term. A Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) thyroid will be calculated at various distances from the plant (site boundary, 2 miles, 5 miles, 10 miles and beyond, if needed). These dose projections are compared to the Protective Action Guides in Procedure AD-EP-ALL-0202, which are derived from the "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents" (EPA 400-R-92-001). Based on these comparisons, protective action recommendations are developed by the Radiological Assessment Manager. If these recommendations involve sheltering, evacuation of the public around the plant or the administration of KI to the public, the Radiological Assessment Manager informs the EOF Director of the situation and recommendations for protective actions.
If dose projections show that PAGs have been exceeded at 10 miles, the dose assessment code and in-field measurements, when available, shall be used to calculate doses at various distances downwind to determine how far from the site PAG levels are exceeded. The Radiological Assessment Manager shall forward the results to the EOF Director who will communicate this information to the off-site authorities.
J.8 Evacuation Time Estimates A description of the methods and assumptions used in developing the analysis of evacuation time estimates is included in the current Evacuation Time Estimate Study for the Catawba Nuclear Site.
(CNS-ETE-12132012, Rev. 000, Part 1 of 2 and Part 2 of 2)
The "evacuation time" is the time between the start of the notification process and the moment the last evacuee crosses out of the area being evacuated. Thus, it includes notification time and time spent preparing to leave, not just travel time.
An updated ETE analysis will be submitted to the NRC under §50.4 no later than 365 days after CNS determination that the criteria for updating the ETE have been met and at least 180 days before using it to form protective action recommendations and providing it to State and local governmental authorities for use in developing offsite protective action strategies.
Rev 148 J-4 December 2019
The criteria for determination that an updated ETE analysis have been met:
- a. The availability of the most recent decennial census data from the U.S. Census Bureau; OR
- a. If at any time during the decennial period, the EPZ permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ to increase by 25 percent or 30 minutes, whichever is less, from the currently NRC approved or updated ETE.
During the years between decennial censuses CNS will estimate EPZ permanent resident population changes once a year, but no later than 365 days from the date of the previous estimate, using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. CNS will maintain these estimates so that they are available for NRC inspection during the period between decennial censuses and shall submit these estimates to the NRC with any updated ETE analysis.
CNS' ETE analysis, using the 2010 decennial census data from the U. S. Census Bureau, was submitted to the NRC via §50.4 on December 13, 2012.
J.9 Implementing Protective Measures If protective actions for any off-site location are deemed necessary, the emergency planning agency of the affected county, in conjunction with the appropriate State agencies (SC-Emergency Planning Division, NC-Department of Crime Control and Public Safety) has the legal authority and responsibility for initiating protective measures for the general public in the plume exposure pathway EPZ including evacuation of these areas. The decision to shelter the public as an alternative to evacuation will be made by the off-site agencies for special populations or when an impediment to evacuation exists. Sheltering in lieu of evacuation should also be considered during a short term release. A short term release is any release that can be accurately projected to be less than the affected protective action zone's evacuation time. An example would be a "puff release." In addition, sheltering may be appropriate (when available) for areas not designated for immediate evacuation because: 1) it positions the public to receive additional instructions; and 2) it may provide protection equal to or greater than evacuation. Public notification of the emergency, the resources used to determine if an evacuation is necessary, the evacuation routes, and the methods used for evacuating persons in the plume exposure pathway EPZ are outlined in the appropriate County and State emergency plans.
For hostile action events, a range of protective actions for onsite workers including evacuation of essential personnel from potential target buildings, timely relocation of non-essential site personnel, dispersal of critical personnel to safe locations, sheltering of personnel away from potential site targets and accountability of personnel after the attack are provided in Emergency Plan Implementing Procedures RP/0/B/5000/026, Site Response to Security Events, RP/0/A/5000/010, Conducting a Site Assembly or Preparing the Site for an Evacuation and AP/0/A/5500/046, Hostile Aircraft Activity.
Rev 148 J-5 December 2019
J.9.a Carowinds: Special Consideration Comprehensive plans provide for early notification to Carowinds of a radiological emergency at Catawba and for evacuation of Carowinds. The plans describe the responsibilities of the emergency response organizations of Mecklenburg and York Counties and provide for the coordination of their efforts among themselves and with Carowinds officials. The plans provide for immediate notification of patrons and staff of Carowinds at the time of the precautionary closing of the park and of the cause of the emergency. Both states and counties located in the ten-mile EPZ agreed that the Charlotte-Mecklenburg Emergency Management Office (CMEMO) will perform the lead planning role regarding a recommended course of action for Carowinds theme park. Refer to Carowinds Standard Operating Procedure (SOP).
See County and State Plans for more detailed information.
J.10 Implementation of Protective Measures for Plume Exposure Pathway J.10.a EPZ Maps Figures i-1 and 2 describe the EPZ's, government jurisdictions, and evacuation zones for Catawba Nuclear Station. Evacuation routes are displayed in Figure J-3.
J.10.b EPZ - Population Distribution Map See Appendix 4, Evacuation Time Estimates.
J.10.c EPZ - Population Alerting and Notification As described in Appendix 3 of this plan, a system exists for alerting and notifying the population (resident and transient) within the EPZ areas. This system is activated by the county and state organization and includes the use of large fixed-site sirens and the Emergency Alert System. A back-up means of alerting and notification is described in the State and County Emergency Plans.
J.10.d EPZ - Protecting Immobile Persons The state and county organization referenced in Section A of this plan have the capability to protect those persons whose mobility may be impaired. The State and County Plans provide for transportation from the person's location to a reception center or shelter.
J.10.e Use of Radioprotective Drugs For Persons in EPZ See State and County plans.
Rev 148 J-6 December 2019
J.10.f Conditions For Use of Radioprotective Drugs See County and State Plans.
J.10.g State/County Relocation Plans See County and State Plans.
J.10.h Relocation Center Locations See County and State Plans.
J.10.i Evacuation Route - Traffic Capacities See County and State Plans.
J.10.j Evacuated Area Access Control See County and State Plans.
J.10.k Planning For Contingencies in Evacuation See County and State Plans.
J.10.l State/County Evacuation Time Estimates The estimates referenced in Appendix 4 are references in the County and State Plans.
J.10.m Bases For Protective Action Recommendations Figure J-2 describes the considerations used by Duke management in developing protective action recommendations.
J.11 Ingestion Pathway Planning See County and State Plans.
J.12 Relocation Center - Registering & Monitoring See County and State Plans Rev 148 J-7 December 2019
DUKE ENERGY CATAWBA NUCLEAR SITE FIGURE J-2, PAGE 1 OF 1 GUIDANCE FOR OFFSITE PROTECTIVE ACTIONS PROTECTIVE ACTION GUIDES (a)
Projected Dose Total Effective Committed Dose Dose Equivalent Equivalent Thyroid (TEDE) (CDE Thyroid) Protective Action Recommendation
< 1 rem < 5 rem No Protective Action is required based on projected dose.
1 rem 5 rem Evacuate affected zones and shelter the remainder of the 10 mile EPZ not evacuated.
N/A 5 rem (b) Consider the use of KI (potassium iodide) in accordance with State Plans and Policy.
(a) Protective Action Guides (PAGs) are levels of radiation dose at which prompt protective actions should be initiated and are based on EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (b) PAG for KI taken from Potassium Iodide as a Thyroid Blocking Agent in Radiation Emergencies, FDA Guidance, November 2001 and Guidance for Industry, KI in Radiation Emergencies, Questions and Answers, FDA, December 2002.
Rev 148 J-8 December 2019
DUKE ENERGY CATAWBA NUCLEAR SITE FIGURE J-3 Catawba Nuclear Station Evacuation Road Network and Nodes Legend 0 Evacuation network nodes with id numbers N Evacuation network
/\./ County boundaries N
D Emergency planning zone boundaries Lakes
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2~"liiiil""'liiii"o~~~2iiiiiiiiiiii..i.4~~~6 Miles J-9 Rev 148 December 2019
INTENTIONAL BLANK PAGE Catawba Nuclear Station Emergency Plan Section P - Responsibility for the Planning Effort P. RESPONSIBILITY FOR THE PLANNING EFFORT To assure that responsibilities for plan development, review and distribution of emergency plans are established and that the Emergency Preparedness Staff is properly trained.
P.1 Emergency Preparedness Staff Training Emergency Preparedness Group personnel attend training/workshops, information exchange meetings with other licensees, and conferences held by industry and government agencies, as available, to maintain current knowledge of the overall planning effort. The Manager, Nuclear Support Services, is required to attend off-site training on an annual basis. This training will be documented in site training files.
P.2 Emergency Response Planning The Site Vice President has the overall authority and responsibility for the Site Emergency Plan.
This planning effort is delegated to the Manager, Nuclear Support Services.
P.3 Manager, Nuclear Support Services The Manager, Nuclear Support Services, has the overall authority and responsibility for radiological emergency response planning as well as the responsibility for the development and updating of the site Emergency Plan and coordination of this plan with other response organizations.
P.4 Review of Emergency Plan Review and updating of the site Emergency Plan shall be certified to be current on an annual basis.
Any changes identified by drills and exercises shall be incorporated into the Site Emergency Plan.
On an annual basis, the Manager, Nuclear Support Services, will provide each state and local organization responsible for off-site activation and protective action decision-making, a copy of the nuclear station procedures appropriate for their area on emergency classification and notification.
A response will be requested by letter within 30 days that a review has been completed of the EALs used for event classification and for protective action recommendations. If problem areas are identified, the Manager, Nuclear Support Services, will ensure resolution.
P.5 Distribution of Revised Plans The Emergency Plan and approved changes shall be forwarded to individuals and organizations listed in Catawba Nuclear Site Document Control Distribution Code CADM-12. Revised pages shall be dated and marked to show where changes have been made. Approved revisions of the site Emergency Plan and Implementing Procedures shall be forwarded to the Site Vice President or designee and made available to the Chairman at the Nuclear Safety Review Board.
Rev. 150 P-1 December 2019
P.6 Supporting Plans Figure P-1 gives a detailed listing of supporting plans to the Catawba Nuclear Station Emergency Plan.
P.7 Implementing Procedures Written procedures will be established, implemented, and maintained covering the activities associated with emergency plan implementation. Each procedure, and changes thereto, shall be reviewed and approved by the responsible implementing manager (line manager or the manager responsible for Emergency Preparedness oversight) prior to implementation and shall be reviewed periodically as set forth in administrative procedures.
Catawba Emergency Plan Implementing Procedures are listed in Figure P-2 with a reference to the section of Emergency Plan implemented by each procedure. Catawba Emergency Plan Implementing Procedures and approved changes shall be forwarded to individuals and organizations listed in Catawba Nuclear Site Document Control Distribution Code CADM-12.
P.8 Table of Contents The Catawba Nuclear Station Emergency Plan contains a specific table of contents. The Catawba Nuclear Station Emergency Plan has been written to facilitate cross-reference to the applicable sections of NUREG-0654 Rev. 1.
P.9 Audit of Emergency Plan The Nuclear Oversight Manager Nuclear QA Audits will arrange for an independent review of Catawba Nuclear Station's Emergency Preparedness Program as necessary, based on an assessment against performance indicators, and as soon as reasonably practicable after a change occurs in personnel, procedures, equipment or facilities that potentially could adversely affect emergency preparedness, but no longer than 12 months after the change. In any case, all elements of the emergency preparedness program will be reviewed at least once every 24 months. Guidance for performing the assessment against the performance indicators is provided in the Emergency Preparedness Administrative Procedure AD-EP-ALL-0001. The independent review will be conducted by the Nuclear Oversight Division which will include the following plans, procedures, training programs, drills/exercises, equipment and state/local government interfaces:
- 1. Catawba Nuclear Station Emergency Plan and Implementing Procedures
- 2. State/Local Support Agency Training Program
- 3. Site Training Program
- 4. Public and Media Training/Awareness
- 5. Equipment - Communications, Monitoring, Meteorological, Public Alerting
- 6. State/Local Plan Interface The review findings will be submitted to the appropriate corporate and nuclear site management.
Appropriate portions of the review findings will be reported to the involved federal, state, and local organizations. The corporate or nuclear site management, as appropriate, will evaluate the findings affecting their area of responsibility and ensure effective corrective actions are taken.
The result of the review, along with recommendations for improvements, will be documented and retained for a period of five years.
Rev. 150 P-2 December 2019
P.10 Telephone Number Updates Telephone numbers listed in the Catawba Nuclear Station Emergency Plan Implementing Procedures will be updated quarterly in accordance with PT/0/B/4600/005B, Quarterly Communications Verification, and TE-EP-ALL-0407, Periodic Verification of EOF Communication Equipment Operation and Equipment/Supply Inventory.
Rev. 150 P-3 December 2019
DUKE ENERGY CATAWBA NUCLEAR STATION FIGURE P-1 SUPPORTING PLANS
- 1. South Carolina Operational Radiological Emergency Response Plan, Appendix 2 South Carolina Emergency Operation Plan (Catawba Nuclear Station, part 4)
- 2. North Carolina Emergency Response Plan for Nuclear Power Facilities (Catawba Nuclear Site, part 2, section 4)
- 3. York County, S.C., Emergency Operations Plan
- 4. Emergency Response Plan, Water Reactors Division, Westinghouse Electric Corporation
- 5. N.R.C. Region II Incident Response Plan
- 6. Interagency Radiological Assistance Plan - Region 3 - U.S. Department of Energy
- 7. INPO Emergency Response Plan Rev. 150 P-4 December 2019
DUKE ENERGY CATAWBA NUCLEAR STATION FIGURE P-2 EMERGENCY PLAN IMPLEMENTING PROCEDURES Procedure # Title Emergency Plan Section Implemented AD-EP-ALL-0100 Emergency Response Section A, B Organization (ERO)
AD-EP-ALL-0101 Emergency Classification Section D, E, I.1 AD-EP-ALL-0103 Activation and Operation of the Section B, C, H Emergency Operations Facility (EOF)
AD-EP-ALL-0104 ERO Common Guidelines and Section B, E, F, G, I, J, K, M Forms AD-EP-ALL-0105 Activation and Operations of the Section B, H, E Technical Support Center (TSC)
AD-EP-CNS-0105 CNS Site Specific TSC Support Section B.H AD-EP-ALL-0106 Activation and Operations of the Section B, H Operations Support Center (OSC)
AD-EP-CNS-0106 CNS Site Specific OSC Support Section B, H AD-EP-ALL-0108 Joint Information System Support Section G AD-EP-ALL-0109 Offsite Protective Action Section J.7 Recommendations AD-EP-ALL-0110 Recovery Section M AD-EP-ALL-0111 Control Room Activation Of The Section D, E, J, K ERO AD-EP-ALL-0202 Emergency Response Off-Site Section D, I, Dose Assessment AD-EP-ALL-0203 Field Monitoring During Declared Section D, I, H.6.b Emergency AD-EP-CNS-0203 CNS Site Specific Field Section I Monitoring Information AD-EP-ALL-0204 Distribution of Potassium Iodide Section J.6 Tablets in the Event of a Radioiodine Release AD-EP-ALL-0205 Emergency Exposure Controls Section K.2 AD-EP-ALL-0301 Activation of the Emergency Section E Response Organization Notification System AD-EP-ALL-0304 State and County Notifications Section E, J.7 AD-EP-ALL-0406 Duke Emergency Management Section F Network (DEMNET)
AD-EP-ALL-0500 Emergency Response Training Section O AD-EP-ALL-0501 Emergency Preparedness Staff Section P Training and Qualification AD-EP-ALL-0801 Design and Development of Drill Section N and Exercises AD-EP-ALL-0802 Conducting Drills and Exercises Section N Rev. 150 P-5 December 2019
Procedure # Title Emergency Plan Section Implemented AD-EP-ALL-0803 Evaluation and Critique of Drills Section N and Exercises AP/0/A/5500/046 Hostile Aircraft Activity Section D RP/0/A/5000/001 Deleted RP/0/A/5000/002 Deleted RP/0/A/5000/003 Deleted RP/0/A/5000/004 Deleted RP/0/A/5000/005 Deleted RP/0/A/5000/006 Deleted RP/0/A/5000/006A Deleted RP/0/A/5000/006B Deleted RP/0/A/5000/006C Deleted RP/0/A/5000/007 Natural Disaster and Earthquake Section D, H, H.6.a RP/0/A/5000/008 Deleted RP/0/B/5000/008 Hazardous Materials Spill Section D
Response
RP/0/A/5000/009 Collision/Explosion Section D RP/0/A/5000/010 Conducting a Site Assembly or Section E, J, K Preparing the Site for an Evacuation RP/0/A/5000/011 Deleted RP/0/B/5000/012 Deleted RP/0/B/5000/013 NRC Notification Requirements Section E RP/0/B/5000/014 Deleted RP/0/B/5000/015 Core Damage Assessment Section D RP/0/B/5000/016 Deleted RP/0/B/5000/017 Deleted RP/0/A/5000/018 Deleted RP/0/B/5000/019 Deleted RP/0/A/5000/020 Deleted RP/0/A/5000/021 Deleted RP/0/B/5000/022 Deleted RP/0/B/5000/023 Deleted RP/0/A/5000/024 Deleted RP/0/B/5000/025 Deleted RP/0/B/5000/026 Site Response to Security Events Section D RP/0/B/5000/028 Deleted HP/0/B/1000/006 Emergency Equipment Functional Section H.10, H.11, K.5 Check and Inventory HP/0/B/1000/010 Determination of Radiation Section D Monitor Setpoints HP/0/B/1009/001 Deleted HP/0/B/1009/003 Deleted HP/0/B/1009/004 Deleted HP/0/B/1009/005 Personnel/Vehicle Monitoring for Section D, J Emergency Conditions Rev. 150 P-6 December 2019
Procedure # Title Emergency Plan Section Implemented HP/0/B/1009/006 Alternative Method for Section D, I Determining Dose Rate within the Reactor Building HP/0/B/1009/007 Inplant Particulate and Iodine Section D, I Monitoring Under Accident Conditions HP/0/B/1009/008 Contamination Control of Injured Section K.5, L.1, L.4 Individuals HP/0/B/1009/009 Deleted HP/0/B/1009/012 Deleted HP/0/B/1009/014 Radiation Protection Actions Section D, I Following an Uncontrolled Release of Liquid Radioactive Material HP/0/B/1009/016 Deleted HP/0/B/1009/017 Deleted HP/(1/2)/B/1009/017 Deleted HP/0/B/1009/018 Deleted HP/0/B/1009/019 Emergency Radio System Section F.1.d Operation, Maintenance and Communication HP/0/B/1009/024 Deleted HP/0/B/1009/025 Deleted HP/0/B/1009/026 Deleted SH/0/B/2005/001 Deleted SH/0/B/2005/002 Deleted SH/0/B/2005/003 Deleted OP/0/A/6200/021 Deleted SR/0/B/2000/002 Deleted SR/0/A/2000/001 Deleted SR/0/A/2000/003 Deleted SR/0/A/2000/004 Deleted Rev. 150 P-7 December 2019