Text

License Amendment Request for Emergency Action Level (EAL) Changes
	ML18247A467
Person / Time
Site:	Callaway
Issue date:	09/04/2018
From:	Wink R; Ameren Missouri, Union Electric Co
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	ULNRC-06433
	Download: ML18247A467 (282)
	v • d • e

,.

Amere MISSOURI n (jllaa Plant September 4. 201$

ULNRC-06433 U. S. Nuclear Regulatory Commission Atm: Document Control Desk Washington. DC 20555-0001 10 CFR 50.47 10 CFR 50.54(q) 10 CFR 50 Appendix F. IV.B.2 10 CFR 50.90 Ladies and Gentlemen:

DOCKET NUMBERS 50-483 AND 72-1045 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

RENEWED FACILITY OPERATING LICENSE NPF-30 LICENSE AMENDMENT REQtJEST FOR EMERGENCY ACTION LEVEL tEAL) CHANGES

References:

1) Regulatory Issue Summary 2003-18. Supplement 2. Use of Nuclear Energy Institute (NE I) 99-01, Methodo logy for Development of Emergency Action Levets. Revision 4 (ADAMS Accession No. MLO5 1450482)

2) NRC Regulatory Issue Summary 2005-02. Revision 1. Clarifying the Process for Making Emergency Plan Changes (ADAMS Accession No. ML100340545)

3) EPFAQ 2016-002. Clarification of Eqtiipment Damage as a Result of a Hazardous Event (ADAMS Accession No. ML17195A299)

4) EPFAQ 2015-0013. Hostile Action resulting in a loss of control of the facility declarations when fuel damage is likely within 4-hours or results in a loss of physical control of spent fuel (ADAMS Accession No. ML16166A366)

Per the guidance provided in References 1. 2. 3 and 4. Union Electric Company (Ameren Missouri) herewith requests a license amendment for changes to Emergency Action Levels (EALs) CA6.1. Cold Shutdown / Refueling S stem Malfunction Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert, and SA9.1. System Malfunction Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert (including addition of a new definition for the term Loss of Safety function (LOSF) and re-definition of the term Visible Damage). along with deletion of Initiating Condition (IC) HGI and associated EAL HGI.l. Hazard HOSTILE ACTION resulting in loss of physical control of the facility: General Emergency, in 83b Counn Road 459 Steedman MO 6S077 ArnerenMissouri corn

U LNRC-06433 September 4. 201 8 Paze 2 of 4 connection with the Callaway Plant Radiological Emergency Response Plan (RERP). These changes.

taken toaether. are conservatively considered to constitute a reduction in effectiveness of the RERP (such that NRC approval is required prior to implementation).

The enclosed package contains Ameren Missouris evaluation of the proposed changes to the EALs and definition, as well as mai-k-ups of the affected EAL Technical Bases Document pages. and a clean copy of the revised EAL Technical Bases Document, as proposed. The evaluation includes the basis for concluding that the Emergency Plan for Callawav Plant will continue to meet the requirements of 10 CFR 50 Appendix E and the planning standards of 10 CFR 50.47 (b). A copy of the proposed EAL matrix (wall charts) is also provided for information.

The Callaway Onsite Review Committee has reviewed and approved the proposed changes and has approved the submittal of this amendment apptication. In addition. in accordance with 10 CFR 50.91 Notice for public comment; State consultation, Section (b)( 1). a copy of this amendment application is being provided to the designated Missouri State official.

Ameren Missouri respectftilly requests approval of this license amendment request by August 30.

2019. Implementation of the revised EALs and EAL Technical Bases Document will be made within 90 days following NRC approval.

This submittal does not contain new commitments. For any questions concerning this letter, please contact Tim Wiltman at 573-590-2225 or Gerry Ranch at 314-225-1727.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely.

Executed on: September 4. 2018 1

__) ,

Roger C. Wink Manager. Regulatory Affairs Attachments: 1) Evaluation of Proposed Emergency Action Levels

2) Mark-ups of Callaway EAL Technical Bases Document Affected Pages

3) EIP-ZZ-00101 Addendum 2, Emergency Action Level Technical Bases Document, Proposed Revision 016 (Clean Copy)

4) EIP-ZZ-0010l Addendum 1, Emergency Action Level Classification Matrix, Proposed Revision 009 (Clean Copy Information Only)

ULNRC-06433 September 4. 2018 Page 3 of 4 cc: Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Reulaton Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. L. John Kios Project Manager. Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 09E3 Washington. DC 20555-0001 Senior Emergency Preparedness Analyst U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011 -45 ii

U LN RC-06433 September 4. 201 8 Page 4 of 4 Index anti send hardcopy to QA file A160.0761 Hardcopy:

Ceiirec Corporation 6100 Western Place. Suite 1050 Fort Worth. TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via RERP ULNRC Distribution:

F. M. Diya T. F. Herrmann S. P. Banker B. L. Cox G. P. Rauch R. C. Wink T. B. Elwood T. R. Wiltman Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. Jay Silbeug (Pillsbury Winthrop Shaw Pittrnan LLP)

Missouri Public Service Commission Mrs. Cotlette Linder. REP Manatzer (SEMA)

Mr. Chris Wiebert (DNR)

Mr. Steve Feeler (DNR)

Mr. Dru Buntin (DNR)

Attachment 1 to ULNRC-06433 Evaluation of Proposed Emergency Action Levels (8 pages)

EvaLuation of Proposed Emergency Action Levels Page 1

1.0 DESCRIPTION

In accordance with the provisions of Part 50, Appendix E, section IV, item B, paragraph 2 and 50.90 of Title 10 of the Code of Federal Regulations(10 CFR), Union Electric Company (Ameren Missouri),

is proposing changes to the emergency action levels tEALs) and their technical bases, as used at Callaway Plant.

Ameren Missouri proposes to modif EALs CA6. 1, CoLd Shutdown / Refueling System Malfunction I lazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert, and SA9. t, System Malfunction Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE: Alert, and eliminate initiating condition (IC) HG! and associated EAL HG! 1, Hazard HOSTILE ACTION resulting in loss of physical control of the facility: General Emergency. Consistent with those changes, Ameren Missouri also proposes to re-define the term VISIBLE DAMAGE and add a new definition for the term LOSS Of SAFETY FUNCTION. Such changes to the EAL scheme require NRC approval prior to implementation. With these changes incorporated, the Emergency Plan for Callaway Plant would continue to meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50.

2.0 PROPOSED CHANGE

A brief descnption of the proposed EAL changes is provided below along with a discussion of the justification for each change.

1. Change the verbiage in the EAL matrix and technical bases for EAL CA6. I which reads:

The occurrence of any Table C-6 hazardous event.

AND EIThER:

Event Damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating MODE.

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating MODE.

to read:

The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EIThER:

Event damage has caused indications of degraded perfomrnnce in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

2. Add Note 11 and Note 12 to the EAL Matrix (in the Notes text box and in the CA6. 1 EAL description text) per Emergency Preparedness Frequently Asked Question (EPFAQ) 20 16-002, Clarification of Equipment Damage as a Result of a Hazardous Event:

Evaluation of Proposed Emergency Action Levels

] Page 2 Note II: 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.

Note 12: This EAL is applicable when a Table C-6, Hazardous Events, causes a LOSS Of SAFETY FUNCTION on a SAFETY SYSTEM required for the current operating MODE, then this emergency classification is not warranted.

3. Change the verbiage in the EAL matrix and technical bases for EAL SA9.l which reads:

The occurrence of any Table S-5 hazardous event.

AND EITHER:

Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating MODE.

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating MODE.

to read:

The occurrence of any Table S-5 hazardous event AND nt damagç has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

4. Add Note 11 and Note 12 to the EAL Matrix (in the Notes text box and in the SA9.

I AL description text) per Emergency Preparedness Frequently Asked Question (EPFAQ) 2016-0 02, Clarification of Equipment Damage as a Result of a Hazardous Event:

Note 11: 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.

Note 12: This EAL is applicable when a Table C-6, Hazardous Events, causes a LOSS OF SAFETY FUNCTION on a SAFETY SYSTEM required for the current operating MODE, then this emergency classification is not warranted.

5. Change the verbiage in the EAL technical bases document definition of VISIBLE DAMA GE, which reads:

Damage to a component or structure that is readily observable without measurement s,

testing or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Evaluation of Proposed Emergency Action Levels Page 3 to read:

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.

6. Add the following verbiage in the PAL technical bases document to provide a definition for the new term LOSS OF SAFETY fLTNCTION:

Loss of Safety Function (LOSf) A LOSF exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be perfornted. for the purpose of the Safety Function Determination Program, a LOSF may exist when a support system is INOPERABLE, AND:

a) A required system redundant to the system supported by the FNOPEFABLE support system is also INOPERABLE; OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR c) A required system redundant to the support system for the supported systems (a) and (b) above is also INOPERABLE.

7. Remove PAL HGI.I and its associated IC HGI from the EAL Matrix per EPFAQ 2015-0013, Hostile Action resulting in a toss of control of the facility declarations when fuel damage is likely within 4-hours or results in a toss of physical control of spent fuel (and shade the resulting empty space in with gray shading to match the other gray shading on the walicharts).

Also remove PAL HG 1.1 and IC I lG I from the EAL Technical Bases Document, and eliminate references to EAL HG 1.1 and IC HGI from the technical bases for other PALs.

Regarding the changes listed as 1, 2, 3, and 4 above, the wording of EALs CA6.l and SA9.l is being revised (and Notes 11 and 12 are being added) in order to preclude an unnecessary Alert declaration when the haiardous event has not caused indication of degraded performance or visible damage on a second train of the same safety system.

Regarding the change listed as 5 above, the definition of VISIBLE DAMAGE is being changed to work more seamlessly with modified EALs CA6.l and SA9.l (including added Note II) described above.

Regarding the change listed as 6 above, a definition of the new term LOSS OF SAFETY FUNCTION as used in modified EALs CA6.l and SA9.1 (including added Note 12) described above is being added to provide clarification.

Regarding the change listed as 7 above, EAL HG 1 I (with its associated IC HG 1)is redundant to other PALs, (i.e., RA2.l, RA2.2, RA2.3, RS2.l, RG2.l, RS1.l, RSI.2, RS1.3, RGI.l, RGI.2, RG1.3, HS 1 .1, 1156.1, HS7, 1, and HG7. 1), which would encompass any escalation from a Hostile Action.

Evaluation of Proposed Emergency Action Levels Page 4 Mark-ups of affected pages and a clean copy of the revised EAL Technical Bases Document are provided as Attachments 2 and 3 to the license amendment request.

3.0 BACKGROUND

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classif, emergency conditions defined in the Callaway Plant Radiological Emergency Response Plan (RERP).

In 2015, the NRC approved use of an EAL scheme for Caltaway Ptant that was developed in accordance with NEt 99-06 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors.

Regarding the NEI 99-01, Revision 6 EALs, the NRC staff position documented in EPFAQ 2016-002 states in part that:

An Alert should be dectared only when actual or potential performance issues with SAFETY SYSTEMS have occurred as a result of a hazardous event. The occurrence of a hazard ous event wilt result in a Notification of Unusual Event (NOUE) classification at a minimum.

In order to warrant escalation to the Alert classification, the hazardous event should cause indications of degraded performance to one train of a SAFETY SYSTEM with either indications of degraded performance on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second SAFETY SYSTEM train, such that the operability or reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoper able, or out of service, prior to the event occurring. As such, the proposed guidance will reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the plant.

It should be noted that the verbiage of the proposed new Note 11 per the changes listed as 2 and 4 above is taken directly from EPFAQ 20 16-002. The proposed new Note 12 per the changes listed as 2 and 4 above, changes to EALs and their technical bases that are listed as 1 and 3 above, and the definitions of the terms VISIBLE DAMAGE and LOSS OF SAFETY FUNCTION per the changes listed as 5 and 6 above provide additional clarification that is consistent with the guidance found in EPFAQ 20 16-002.

Regarding the change listed as 7 above, Ameren Missouri originally developed IC HG I and EAL HG1.l in accordance with NEt 99-01, Revision 6. Since implementation, many questions have arisen that could not be clearly answered due to the wording of the IC and EAL. further more, it has been noted that if a Hostile Action were to occur and major safety functions were lost (or damage to the fuel pool occurred), any accident that could cause a concern for the health and safety of the public would be identified by one or more of the othe; existing EALs, In regards to this issue, the NRC staff position documented in EPFAQ 2015-0 13 states in part that:

Consideration can be given to not include EAL HGI in a site-specific EAL scheme.

However, EALs AA2. AS2, AG2, ASI, AG1, HSI, HS6, HS7, and HG7 shall be as provided in Nil 99-01, Revision 6

Evaluation of Proposed Emergency Action Levels Page 5 Consistent with this guidance. Ameren Missouri has developed and implemented NE! 99-01 Revision 6 ICs AA2, AS2, AG2, AS1, AG!, HSI, H56, HS7, and 11G7, as endorsed by the NRC. In the Callaway RERP, the equivalent lCs are RA2, RS2, RG2, RSI, RGI, 1151, 11S6, HS7, and HG?,

respectively. EAL HG1.l (and its associated IC HG1), which is proposed to be removed from the Callaway RERP, is redundant to EALs RA2.1, RA2.2, RA2.3, RS2.1, RG2.I, RS.1.l, RSI.2, RSI.3, RG1.l, RG1.2, RGI.3, HS1.l, 1156.1, HS7.1, and HG?.!, as these EALs would encompass any escalation from a Hostile Action.

4.0 TECHNICAL ANALYSIS

The proposed changes affect the Callaway Plant Radiological Emergency Response Plan (RERP) and otherwise do not alter requirements of the Operating License or the Technical Specifications. These changes do not alter any of the assumptions used in the safety analyses, nor do they cause any safety system parameters to exceed their acceptance limits. Therefore, the proposed changes have no adverse effect on plant safety.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements! Criteria Per 10 CFR 50.54(q)(2):

A holder of a license under this part, or a combined license under part 52 of this chapter after the Commission makes the finding under § 52.103(g) of this chapter, shall follow and maintain the effectiveness of an emergency plan that meets the requirements in appendix F. to this part and, for nuclear power reactor licensees, the planning standards of § 50.47(b).

10 Cf R 50.47(b)(4) requires the emergency response plan to meet the following standard:

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

10 CfR 50.54(q)(4) states:

The changes to a licensees emergency plan that reduce the effectiveness of the plan as defined in paragraph (q)( 1 )(iv) of this section may not be implemented without prior approval by the NRC. A licensee desiring to make such a change after february 21, 2012 shall submit an application for an amendment to its license. In addition to the filing requirements of 50.90 and 50.91, the request must include all emergency plan pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the licensees emergency plan, as revised, will continue to meet the requirements in appendix F. to this part and, for nuclear power reactor licensees, the planning standards of 50.47(b).

Evaluation of Proposed Emergency Action Levels Page 6 As defined in 10 CFR 50.54(q)(l)(iv):

Reduction in e/fectheness means a change in an emergency plan that results in reducing the licensees capability to perform an emergency planning function in the event of a radiological emergency.

The proposed changes listed as 1,2,3,4,5, and 6 above are consistent wIth the NRC staff position documented in EPFAQ 2016-002. and Ameren Missouri has therefore concluded that the Emergency Plan for Callaway Plant would continue to meet the planning standards of 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. However, a classification based on NRC-endorsed industry guidance in NEI 99-0 1, Revisions 4, 5, and 6, as well as in NUMARCINESP-007, could be different from a classification based on the revised EALs. Therefore, this proposed change is considered a deviation in accordance with Regulatory Issue Summary (RIS) 2003-18, Supplement 2,

Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels, Revision 4. Deviations are considered to be a reduction in effectiveness, and thus, prior NRC approval is required.

The proposed change tisted as 7 above is consistent with the NRC staff position documented in EPFAQ 2015-0 13, and Ameren Missouri has therefore concluded that the Emergency Plan for Callaway Plant would continue to meet the planning standards of 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. However, this proposed change is also considered a

deviation in accordance with Regulatory RIS 2003-1 Supplement 2, and thus, prior NRC approval is required.

5.2 No Significant Hazards Consideration Ameren Missouri has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No, The proposed changes to the Callaway Plant emergency action levels do not impact the physical function of plant structures, systems, or components (SSC) or the maimer in which SSCs perform their design function. The proposed changes have no effect on accident initiators or precursors, nor do they alter design assumptions. The proposed changes do not alter prevent or the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Evaluation of Proposed Emergency Action Levels Page 7

2. Does the proposed amendment create the possibility of a new or different kind of accident jvnz any accident previously evaluated?

Response: No. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed, and no equipment will be removed), nor do the proposed changes involve a change in the method of plant operation. The proposed changes will not introduce failure modes that could result in a new accident, nor do the changes alter assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a sign flcant reduction in a ,nargin of safety?

Response: No. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. There are no changes to setpoints or environmental conditions of any SSC or the mariner in which any SSC is operated. Margins of safety are unaffected by the proposed changes. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

Therefore, the proposed changes do not involve any reduction in a margin of safety.

Based on the above, Ameren Missouri concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and accordingly, a finding of no significant hazards consideration is justified.

6.0 ENVIRONMENTAL CONSIDERATION

S The proposed changes to the emergency action levels maintain the environmental bounds of the current environmental assessment associated with the Callaway Plant Unit 1. The proposed changes will not affect plant safety and will not have an adverse effect on the probability of an accident occurring. The proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a siimiflcant increase in individual or cumulative occupational radiation exposure. Therefore, no enviromnental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 PRECEDENTS The proposed changes listed as 1, 2, 3, 4, 5, and 6 above are similar in nature to changes that were included in a license amendment request (LAR) for changes to Monticello Nuclear Generating Plant EALs that was submitted by Northern States Power Company (Xcel Energy) on September 27, 2017 (ADAMS Accession No. ML17269A076).

The proposed changes listed as 7 above are similar in nature to changes that were included in a LAR for changes to the Beaver Valley Power Station EALs that was submitted by First Energy Nuclear Operating Company on August 10, 2017 (ADAMS Accession No. ML17222A219).

Ameren Missouri is not aware of any similar LARs that have yet been approved by the NRC.

Evaluation of Proposed Emergency Action Levels Page 8

8.0 REFERENCES

1. EPFAQ 20 16-002, Clarification of Equipment Damage as a Result of a Hazardous Event (ADAMS Accession No. NIL 171 95A299)

2. EPFAQ 20 15-013, Hostile Action resulting in a loss of control of the facility declara tions when fuel damage is likely within 4-hours or results in a loss of physical control of spent fuel (ADAMS Accession No. ML16166A366)

3. NRC Regulatory issue Summary 2005-02, Revision 1, Clarifying the Process for Making Emergency Plan Changes (ADAMS Accession No. MLI 00340545)

4. NEI 99-01, Revision 6, ivlethodology for the Development of Emergency Action Levels for Non-Passive Reactors (ADAMS Accession No. ML12326A805)

5. Regulatory Issue Summary 2003-18, Supplement 2, Use of Nuclear Energy institu te (NE 1) 99-01, Methodology for Development of Emergency Action Levels, Revision 4 (ADAMS Accession No. MLO5 1450482)

6. NUMARC/NESP-007. Rev. 2, Methodology for Development of Emergency Action Levels (ADAMS Accession No. ML041120174)

Attachment 2 to ULNRC-06433 Mark-ups of Callaway EAL Technical Bases Document Affected Pages (19 pages)

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 reqtiired if large quantities of smoke and heat are observed.

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

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

General Emergency Events are in process 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.

High Winds Winds in excess of 40 mph (18 mIs) sustained, or 58 mph (26 m/s) gusting.

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

hostile Action An act toward Callaway 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 destrtictive 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 Callaway. 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 deteriuinecl 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 eqclipment. such as SCBAs, that is not routinely employed).

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

Initiating Condition An event or condition that aligns with the definition of one of the four emergency classification tevets by virtue of the potential or actual effects or consequences.

Page 20 of 241 INFORMATION USE

\#1here]

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Site Area Emergency Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile actions that result in intentional damage or malicious acts: (I) 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 Guidelines exposure levels beyond the site boundary.

Site Boundary Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1.200 meters (3,937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with IOCFRIOO (ref. 4.1.12).

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

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

Valid An indication, report, or condition. is considered to be valid when it is verified by l) 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.

Page 22 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 6.0 CALLA WAY-TO-NFl 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Callaway EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Caltaway EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

Callaway NEI 99-0 1 Rev. 6 Callaway NEI 99-01 Rev. 6 Callaway NEI 99-01 Rev. 6 Example Example Example EAL IC EAL IC EAL IC RU1.1 AU1 1,2 CU5.i CU5 1,2,3 HA5.1 HA5 RU1.2 AU1 3 CA1.1 CAl 1 HA6.1 HA6 RU2.1 AU2 1 CA1.2 CAl 2 HA7.1 HA7 RA1.1 AA1 1 CA2.1 CA2 1 HS1.l HS1 1 RA1.2 AA1 2 CA3.l CA3 1,2 HS6.l HS6 1 RA1.3 AA1 3 CA6.l CA6 1 Hi-- RA1.4 AA1 4 cSi.i csi 1 HG1.1 -He-I RA2.1 AA2 1 CS1.2 CS1 2 H- G ---

RA2.2 AA2 2 CS1.3 CS1 3 SU1.1 SU1 1 RA2.3 AA2 3 CG1.i CG1 1 SU3.l SU2 1 RA3.1 AA3 1 001.2 CG1 2 SU4.l SU3 2 RA3.2 AA3 2 FA1.1 FA1 1 SU5.l SU4 1,2,3 RS1.1 AS1 1 FS1.l FS1 1 SU6.l SU5 1 RS1.2 AS1 2 FG1.1 FG1 1 SU6.2 SU5 2 RS1.3 AS1 3 HU1.1 HU1 1,23 5U7.1 SU6 1,2,3 RS2.1 AS2 1 HU2.i HU2 1 SU8.i SU7 1,2 RG1.1 AG1 1 HU3.i HU3 1 SA1.1 SAl 1 RG1.2 AG1 2 HU3.2 HU3 2 SA3.l SA2 1 RG1.3 AOl 3 HU3.3 HU3 3 SA6.l SA5 1 RG2.1 AG2 1 Hu3.4 HU3 4 SA9.l SA9 1 CUJ.1 Gui 1 HU4.i HU4 1 551.1 551 1 CU1.2 cui 2 Hu4.2 HU4 2 SS2.l SS8 1 CU2.1 CU2 1 HU4.3 HU4 3 SS6.1 SS5 CU3.1 CU3 1 HU4.4 HU4 4 SG1.l 501 CU3.2 CU3 2 HU7.1 HU7 1 501.2 SG8 1 CU4.i CU4 1 HA1.i HAl 1,2 EU1.1 E-HUJ 1 Page 25 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 RUI. is tised here to distinguish between non-emergency anti 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 e e i e e r 1 e o i a ir t f c s r e t - n St n from the cask.

Security-related events for ISFSIs are covered tinder 1G-sH-G1and HSI. IC HS1

1. Certificate of Compliance No. 1040 Appendix A Technical Specifications for the Fit-STORM UMAX Canister Storage System

2. NEI 99-01. E-HUI Page 69 of 241 INFORMATION USE

EIP-ZZ-OOtOl ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Matfunction Subcategory: 6 Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE Table C-6 Hazardous Events

[Insert #3 here] :

EXPLOSION HIGH WINDS or tornado strike

Internal or external FLOODING event

Seismic event (earthquake)

Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

EXPLOSION A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or over pressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circtiits, grounding, arcing, etc.) should not autotiiatically 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.

HIGH WINDS Winds in excess of 40 mph (18 mIs) sustained, or 58 mph (26 mIs) gusting.

Page 111 of 241 INFORMATION USE

here) EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases SAFETY SYSTEM A system required for sate plant operation, cooling down the plant and/or placing it in the Cold Shutdown condition. including the ECCS. These are typically systems classified as safetyrelated (as defined in 10CFR5t).2):

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

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 V g to a component or structure that is readily observable witho testing. or analysis. The visual im act . se concern regarding the operability

Annunciator 98D. OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-0000l, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

Internal FLOODING may be catised by events such as component failures, eqtiipment misalignment.

or outage activity mishaps (ref. 2).

External flooding may he dime to high rainfall. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph. (ref. 4).

Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 5).

An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a enough to cause concern regarding the operability ility of the SAFETY SYSTEM train.

The second conditional addresses damage AFETY SY component that is not in service/operation or readily apparent through indicj, alone, or to a structure contain ,.SAFETY SYSTEM components.

Operators will make this trrff{nation based on the totality of available evt-aQdamage report information. ThiJs_iiiended to be a brief assessment not requiring lengthy analysibqantification of the damage.

of the emergency classification level would be via IC CS I or RS 1.

Eelon

\

\((iTisert #6 here]

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. OTO-SG-00001, Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR, Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF. Groundwater, Reference, and Acttial Plant Elevations

4. FSAR, Section 3.3.1.1 Design Wind Loadings

5. FSAR, Section 9.5. 1 Fire Protection System

6. NEI 99-0 1, CA6 Page 113 of 241 INFOR1/IATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Eniergency Action Level Technical Bases Category: H Hazards Subcategory: 1 Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

FHJ1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shill Supervisor.

OR Notification of a credible security threat directed at the site.

OR A validated notification from the NRC providing information of an au-craft threat.

MODE Applicability:

All Definition(s):

SECURITY CONDITION Any sectirity 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 ACT/ON An act toward Callaway 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 detiver destructive force.

Other acts that satisfy the overall intent may be included. Hostile action shotild not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Callaway.

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 security shift stipervision is defined as the Security Shift Supervisor.

This EAL is based on the Callaway Plant Security Plan and DBT (ref. I).

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 . o e n nts assessed as HOSTILE ACTIONS are classifiable tinde ICs HA-1,--HS 1 a4-gG-l HAl and HS1.

Timely and accurate communications between Security i p v io a n o oom 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.

Page 115 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical 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 Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement ptiblic protective actions in the unlikely event that the attack is successful in impairing multiple safety fttnctions.

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 httnters, 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, stich as the partictilars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan and DBT Escalation of the emergency classification level would be via IG-HG-L Cs RG1, RG2, SG1, SG2, FG1, and CG

1. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOO1, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-0000 1, Plant Security Event Hostile Intrusion

5. OTO-SK-00002, Plant Security Event - Aircraft Threat

6. NE199-01. HSI Page 120 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM/

Rev .,0l5 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT

/

Attachment 1 Emergency Action Level Technical Bases

- //

HOSTILE ACTION resulting in loss of physical control of the f

/

ility HG1J neral Emergency A HOSTILE ACTION\occurring or has occurred within the PROTECT AREA as reported by the Security Shift Supervi\r.

AND EITHER of the fol1ow\g has occurred:

Any of the following ety functions cannot be controll or maintained Reactivity control.

Core cooling.

RCS heat removal.

OR

Damage to spent fuel has occurred is IM INENT.

MODE Applicability:

All Definition(s):

HOSTILE ACTION An act toward Calla ay or its personnel tha inclcides the cisc of violent force to destroy equipment. take hostages. and/or intimi ate the licensee to achieve i end. This includes attack by air, land.

or water using guns, explosives. proje lIes, vehicles, or other devices ised to deliver destructive force.

Other acts that satisfy the overall in nt may be iticluded. Hostile action hould not be construed to include acts of civil disobedience or feb otis acts that are not part of a concerted ttack on Callaway.

Non-terrorism-based EALs sho d be used to address such activities (i.e.. thi, may include violent acts between individuals in the o er controlled area).

IMMINENT- The trajecto of events or conditions is such that an EAL will be within a relatively short period of time regardles of mitigation or corrective actions.

PROTECTED AREA An area encompassed by physical barriers and to which access i controlled. The Protected Area ref s to the designated security area around the process buildings and is

$600-X-88 100 P operty-Site Layout. Owner Controlled Area and Surrounding Area.

7 Page 121 of 241 INFORMATION USE

flNCNNCNcN-EIP-ZZ-00 101 ADDENDUM%

EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Basis:

The security s lit supervisioti is deflned as the Security Shift Supervisor.

This IC addresse an event in which a HOSTILE FORCE has taken physical control of the acility to the extent that the plan staff can no longer operate equipment necessary to maintain key safe functions. It also addresses a HOSTIL ACTION leading to a loss of physical control that results in actt I or IMMINENT damage to spent fuel dt to t) damage to a spent fuel pool cooling system (e.g., pu s. heat exchangers, controls, etc.) or, 2) loss spent ftiel pool integrity such that scifficient water leve annot be maintained.

Timely and accurate commu cations between Security Shift Stipervision and t Control Room is essential for proper classification of a se urity-related event (ref. 2, 3).

Security plans and terminology ar based on the guidance provided by NE 03-12, Template for the Security Plan, Training and Qualification Pla Safeguards Contingency Plan ant ndependent Spent Fuel Storage Installation Security Program.

Emergency plans and implementing proc tires are pttblic docume ts; therefore, EALs should not incorporate Security-sensitive information, his inclttdes infor tion that may be advantageocis to a potential adversary, such as the particcdars cot erning a specif threat or threat location. Security-sensitive information should be contained in non-public curnents s h as the Callaway Plant Sectirity Plan & DBT (ref.l).

Callaway Basis Reference(s):

I. Callaway Plant Sectirity Plan and DBT (Safc

2. EIP-ZZ-SKOOI, Response to Sectirity

3. SDP-CP-00003, Security Contingency4nts

4. OTO-SK-0000l, Plant Security Eve/Hostile Intrusion

5. OTO-SK-00002, Plant Security Ønt Aircraft Threat

6. NEI 99-01. HGI Page 122 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 9 Hazardous Event Affecting Safety Systems Initiating Condition: Hazardotis event affecting a SAFETY SYSTEM needed for the current operating MODE tiiieteeded for the current operating MODE.

Table S-5 Hazardous Events Insert #7 here]

EXPLOSION

FIRE

HIGH WINDS or tornado strike

Internal or external FLOODING event

Seismic event (earthquake)

Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

- Power Operation. 2 - Starttip. 3 - Hot Standby. 4 - Hot Shtttclown Definition(s):

EXPLOSION A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or over pressurization. A release of steam (from high energy hues 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.

HIGH WiNDS Winds in excess of 40 mph (1$ mis) sustained, or 58 mph (26 mIs) gusting.

Page 18$ of 241 INFORMATION USE

I[1nse #8 here] EIP-ZZ-00101 ADDENDUM 2 Rev. 015

/ EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases SAFETY SYSTEM A system required for safe plant operation, cooling down the pLant and/or placing it in the Cold Shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR5O.2):

Those strttctures. systems and components that are relied upon to remain functional during and following design basis events to asstire:

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 cotild result in potential VI to a component or structure that is readily observabl -

testing. or analysis. The visual im act of se concern regarding the operability Basis: f#9here1

Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the 031 threshold. OTO-SG-0000 1, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

Internal F10ODING may be caused by events such as component failures, eqtlipment misalignment.

or outage activity mishaps (ref. 2).

External flooding may be due to high lake level. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

Seismic Category I structures are analyzed to withstand a stistained. design wind velocity of at least 100 mph. (ref. 4).

Areas containing functions and systems required for safe shtitdown of the plant are identified by fire area (ref. 5).

An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.

A single FAULTED steam generator would NOT reqclire declaration per this EAL. Technical Specification Bases 3.7.4 explains that two intact Steam Generators are required for cooldown of the RCS and a third Steam Generator is assumed to be RUPTURED. If more than one Steam Generator is FAULTED, then this Page 189 of 241 INFORMATION USE

\j[lnsert #10 here]

EtP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT inforination. This is intended to be a brief nt no lengthy anal sis or quantification of the -

damage. -

ion of the emergency classification level would be via IC FS I or RS 1.

Callaway Basis Reference(s):

1. OTO-SG-00001. Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR. Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF. Groundwater, Reference, and Actual Plant Elevations

4. FSAR. Section 3.3.1.1 Design Wind Loadings

5. FSAR, Sectioti 9.5.1 Fire Protection System 6, NFl 99-01. SA9 Page 190 of 241 INFORMATION USE

Insert #1:

Loss of Safety Function (LOSF) A LOSF exists when, assttming no conculTent single failure, a safety function assumed in the accident analysis cannot be performed. for the purpose of the Safety Ftinction Determination Program, a LOSF may exist when a support system is flJOPERABLE. AND:

a) A required system redundant to the system supported by the INOPERABLE support system is also [NOPERABLE OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also fNOPERABLE; OR c) A required system redundant to the support system for the supported systems (a) and (b) above is also INOPERABLE.

Insert #2 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.

Insert #3 CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(Votes II. 12)

Vole 1?: If the hazardous event only results in VISIBLE DAMAGE, with no indication of degraded performance to at least one train ofa SAFETY SYSTEM, then this emergency classification is not warranted.

Vole 12: This EAL is applicable when a Table C-6, 1-lazardous Events, causes a LOSS OF SAFETY FUNCTION on a SAFETY SYSTEM requited fot the current operating MODE.

Insert #4 LOSS OF SAFETY FUNCTION (LOSF) A LOSF exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of the Safety Function Determination Program, a LOSF may exist when a support system is INOPERABLE, AND:

a) A required system redundant to the system supported by the INOPERABLE support system is also INOPERABLE; OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR c) A required system redundant to the support system for the supported systems (a) and (b) above is also INOPERABLE.

Insert #5 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.

Insert 6 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; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via Cs CS1 or RS1.

Insert #7 SA9.1 Alert The occurrence of any Table 5-5 hazardous event AND Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(Votes II, /2)

Vote II: Ifthe hazardous event only results in VISIBLE DAMAGE, with no indication of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Vote /2: This EAL is applicable when a Table S-5 Hazardous Events causes a LOSS Of SAFETY FUNCTION on a SAFETY SYSTEM reqtured tor the current operating MODE.

Insert #8 LOSS OFSA FETY FUNCTION (LOSF) A LOSF exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of the Safety Function Determination Program, a LOSF may exist when a support system is INOPERABLE, AND:

a) A required system redundant to the system supported by the INOPERABLE support system is also INOPERABLE; OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR C) A required system redundant to the support system for the supported systems (a) and (b) above is also INOPERABLE.

Insert #9 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.

Insert #10 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, commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via ICs FS1 or RS1.

Attachment 3 to ULNRC-06433 HP-ZZ-OO 101 Addendum 2, Emergency Action Level Technical Bases Document, Proposed Revision 016 (Clean Copy)

(239 pages)

AmerenMISSOURI Cailaway Energy Center EIP-ZZ-OO1O1 ADDENDUM 2 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT MAJOR Revision 016 Page 1 of 239 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number 1.0 PURPOSE 6 2.0 DISCUSSION 6 2.1. Background 6 2.2. fission Product Barriers 7 2.3. fission Product Barrier Classification Criteria 7 2.4. EAL Organization 8 2.5. Technical Bases Information 10 2.6. Operating MODE Applicability (ref. 4.1.8) 11 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 12 3.1. General Considerations 12 3.1.1. Classification Timetiness 12 3.1.2. Valid Indications 12 3.1.3. Imminent Conditions 13 3.1.4. Planned vs. Unplanned Events 13 3.1.5. Classification Based on Analysis 13 3.1.6. Emergency Coordinator Jtidgment 13 3.2. Classification Methodology 14 3.2.1. Classification of Multiple Events and Conditions 14 3.2.2. Consideration of MODE Changes During Classification 14 3.2.3. Classification of Imminent Conditions 15 3.2.4. Emergency Classification Level Upgrading and Downgrading 15 3.2.5. Classification of Short-Lived Events 15 3.2.6. Classification of Transient Conditions 16 3.2.7. After-the-Fact Discovery of an Emergency Event or Condition 17 3.2.8. Retraction of an Emergency Declaration 17

4.0 REFERENCES

1$

4.1. Developmental 18 4.2. Implementing 18 5.0 DEFINITIONS, ACRONYMS, & ABBREVIATIONS 19 5.1. Definitions (ref 4.1.1 except as noted) 19 5.2. Abbreviations/Acronyms 23 6.0 CALLAWAY-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE 25 7.0 ATTACHMENTS 26 Page 2 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number 8.0

SUMMARY

OF CHANGES 26 ATTACHMENT 1 Emergency Action Level Technical Bases 31 Category R Abnormal Rad Release / Rad Effluent 31 RUI.1 Untisual Event 32 RU 1.2 Unusual Event 34 RAY.1 Alert 36 RA1.2 Alert 3$

RAI.3 Alert 40 RAI.4 Alert 42 RS1.1 Site Area Emergency 44 RSI.2 Site Area Emergency 46 RS1.3 Site Area Emergency 48 RG 1 .1 General Emergency 50 RG1 .2 General Emergency 52 RG1.3 General Emergency 54 RU2.l Unusual Event 56 RA2.1 Alert 58 RA2.2 Alert 59 RA2.3 Alert 61 RS2.1 Site Area Emergency 62 RG2.1 General Emergency 63 RA3.l Alert 64 RA3.2 Alert 65 Category E Independent Spent Fuel Storage Installation (ISfSI) 67 EUI.1 Unusual Event 6$

Category C Cold Shutdown / Refueling System Malfunction 70 CU 1.1 Unusual Event 71 CUI.2 Unusual Event 73 CA1.l Alert 75 CAI.2 Alert 77 CS 1.1 Site Area Emergency 79 CS 1.2 Site Area Emergency 81 CS 1.3 Site Area Emergency $3 CG1.1 General Emergency $6 CG1.2 General Emergency 90 CU2.1 Unusual Event 94 CA2.l Alert 97 CU3.l Unusual Event 99 Page 3 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number CU3.2 Unusual Event 101 CA3.1 Alert 103 CU4.1 Unusual Event 106 CU5.l Unusual Event 108 CA6. 1 Alert 111 Category H Hazards and Other Conditions Affecting Plant Safety 114 HUI.l Unusual Event 115 KAI.1 Alert 117 HSI.1 Site Area Emergency 119 HU2.1 Unusual Event 121 HU3.1 Unusual Event 123 HU3.2 Unusual Event 124 HU3.3 Unusual Event 126 HU3.4 Unusual Event 127 HU4.1 Unttsual Event 128 HU4.2 Untisual Event 131 HU4.3 Unusual Event 134 HU4.4 Unusual Event 135 HA5.1 Alert 136 HA6.1 Alert 138 HS6.1 Site Area Emergency 139 HU7.1 Unusual Event 141 HA7.1 Alert 142 HS7.1 Site Area Emergency 144 HG7.1 General Emergency 146 Category S System Malfunction 148 SUI.1 Unusual Event 150 SAI.1 Alert 152 SS1.1 Site Area Emergency 155 SG1.1 General Emergency 157 SGI.2 General Emergency 159 SS2.1 Site Area Emergency 162 SU3.1 Unusual Event 164 SA3.l Alert 166 SU4.1 UnusualEvent 168 SU5.1 Unusual Event 169 SU6.l Unusual Event 171 SU6.2 Unusual Event 174 SA6.1 Alert 177 Page 4 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number SS6.1 Site Area Emergency 179 SU7. 1 Unusual Event 1 81 SU8. 1 Unusual Event 184 SA9.1 Alert 186 Category F fission Product BalTier Degradation 189 FAI.l Alert 191 fS1.1 Site Area Emergency 192 FGI.1 General Emergency 193 ATTACHMENT 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases 194 Page 5 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL).

Decision-makers responsible for implementation of EIP-ZZ-00101, Classification of Emergencies, should use this document as a technical reference in support of EAL interpretation. This infot-mation may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be usefttl in training and for explaining event classifications to offsite 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).

Additionally, changes to plant AOPs and EOPs that may impact EAL bases shall be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION 2.1. Backuround EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the Callaway Plant Radiological Emergency Response Plan (RERP).

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EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 prodticts 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 balTier loss and decreased certainty of maintaining the barrier.

The primary fission product bathers are:

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

B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections tip to and including the primary isolation valves.

C. Containment (CMI): The Containment Barrier includes the containment buildintt 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.

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

Alert:

Any loss or any potentiaL loss of either Fuel Clad or RCS barrier Site Area Emerencv:

Loss or potential loss of any two barriers General Emerencv:

Loss of any two barriers and loss or potential loss of the third baiTier Page 7 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.4, EAL Organization The Catlaway EAL scheme includes the following features:

Division of the EAL set into three broad groups:

EALs applicable under all conditions This group would be reviewed by the EAL-user any time emergency classification is considered.

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, Startttp, or Power Operation MODE.

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 Callaway EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the Callaway scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The Callaway EAL categories and stihcategories are listed below.

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EIP-ZZ-0t)101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT EAL Grotips, Categories and Subcategories EAL Group I Category EAL Subcategory All Conditions:

R Abnormal Rad Levels / Rad Effluent 1 Radiological Effluent 2 Irradiated Fuel Event 3 Area Radiation Levels I ESFSI I Confinement Boundary H Hazards and Other Conditions Affecting 1 Security Plant Safety 2 Seismic Event 3 Natural or Technological Hazard 4 Fire 5 Hazaidotis Gases 6 Control Room Evacuation 7 Emergency Coordinator Judgment Hot Conditions:

S System Malfunction 1 Loss of Emergency AC Power 2 Loss of Vital DC Power 3 Loss of Control Room Indications 4 RCS Activity 5 RCS Leakage 6 RTS Failure 7 Loss of Communications 8 Containment Isolation Failure 9 Hazardous Event Affecting Safety Systems -

F Fission Product Barrier Degradation None Cold Conditions:

C Cold Shutdown / Reftteling System I RCS Level Malfunction 2 Loss of Emergency AC Power 3 RCS 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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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.5. Technical Bases Information EAL technical bases are provided in Attachment I for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, S, E and F) 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 Subcategorv Number & Title Initiating Condition (IC)

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

EAL Identifier (enclosed in rectaiigle)

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 identi tier:

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

2. Second character (letter): The emergency classification (G, 5, A or U)

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

3. Third character (number): Subcategoiy 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 numericat sequence of the EAL within the EAL subcategory. If the subcategoty has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (5) or General Emergency (G).

F AL (enclosed in rectangle)

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

MODE Applicability One or more of the following plant operating conditions comprise the MODE to which each EAL is applicable: I Power Operation, 2 Startup, 3 Hot Standby, 4 Hot Shutdown, 5 Cold Shutdown, 6 Refueling, D Deftieled, or Any. (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.

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EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Basis:

A Plant-Specific basis section that provides Callaway-relevant information concerning the EAL. This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NFl 99-0 1 Rev. 6.

Catlawav Basis Reference(s):

Site-specific source documentation from which the EAL is derived.

2.6. Operating MODE Applicability (ref. 4.1.8)

I Power Operation Keii? 0.99 and reactor thermaL power> 5%.

2 Startup Kei? 0.99 and reactor thermal power 5%.

3 Hot Standby KeH <0.99 and average coolant temperature? 350°F.

4 Hot Shutdown Kti 0.99 and average coolant temperat-ure 350°F

< > T> 200°F and at least 53 of 54 reactor vessel head closure bolts fully tensioned.

5 Cold Shutdown Kerr < 0.99 and average coolant temperature 200°F.

6 Refueling Two or more reactor vesseL head closure bolts are less than frilly tensioned.

D Defueled All fttel assemblies have been removed from Containment and placed in the spent fuel pit and the SFP transfer canal gate valve is closed.

The MODE in etTect 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.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 toss of fission Product Barrier Thresholds.

3.1.1. Classification Timeliness NRC regulations require the licensee to estabtish 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 ctassification level. The NRC staff has provided guidance on implementing this requirement in NSIRDPR-ISG-O 1, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants (ref. 4.1.11).

Time based EALs should be evaluated upon first indication of the conditions. If someone is working to mitigate the condition in less than the time required, the declaration can wait to see if they are successful within the time constraints. If there is indication that the threshold will be exceeded for the time period, the declaration should immediately be declared, regardless of the time remaining. In the case of leaks, the exceeded threshold will take some additional period of time to lower and must he taken into accotint.

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.

3.1.2. Valid Indications All emergency classification assessments shall be based upon vatid 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. The validation of indications shoutd be completed in a manner that supports timely emergency declaration.

An indication, report, or condition is considered to be valid when it is verified by (I) 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.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 tintil 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 he assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4. Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an PAL 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. [n these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.4).

3.1.5. Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis disctission will identify the necessary analysis. In these cases, the 1 5-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 infonnation 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 Judurnent 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 ji.idgment is still necessary. The NEI 99-01 PAL 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.

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EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 15 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 NSIRIDPR-ISG-01 (ref. 4.1.11).

3.2.1. Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared and the Alert noted in facility logs.

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

If two Alert EALs are met, one of the Alerts should be declared and the other Alert should be noted in the facility logs.

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 he 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.

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EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 alt emergency classification levels, this approach is particularly important at the higher emergency classification Levels since it provides additional time for implementation of protective measures.

3.2.4. Emergency Classification Level Upradin and Down%radinu 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 he 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 cotipleted. 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 tril) the reactor followed by a successful manual trip.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.2.6. Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will reqtiire 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.

E AL Momentarily Met During Expected Plant Response In instances where an EAL is briefly met during an expected (nonnal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performecL 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 RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should he based on the ATWS only.

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses onLy those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency 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.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 32.7. 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.

3.2.8. Retraction of an Emeruency 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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EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT

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 Qtiicltly Changing Events, February 2, 2007.

4.1.3. NUREG-1022 Event Reporting Guidelines: IOCfR5O.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. Drawing $600-X-8$ 100 Property-Site Layout Owner Controlled Area and Surrounding Area 4.1.7. Callaway ESAR Figure 1.2-44 Plant Area Layout 4.1.8. Technical Specifications Table 1.1-1 MODES 4.1.9. OSP-GT-00003 Containment Closure 4.1.10. Procedure Writers Manual Callaway Plant Procedure Writers Manual 4.1.11. NSIR/DPR-ISG-0l Interim Staff Gtiidance, Emergency Planning for Nuclear Power Plants 4.1.12. Callaway Plant Radiological Emergency Response Plan Emergency Plan (RERP) 4.1.13. OTG-ZZ-00007 Refueling Preparation, Performance and Recovery 4.1.14. APA-ZZ-00520, Reporting Requirements and Responsibilities 4.2. Implementing 4.2.1. EIP-ZZ-00 101 Classification of Emergencies 4.2.2. NEI 99-01 Rev. 6 to Callaway EAL Comparison Matrix 4.2.3. Callaway EAL Matrix 4.2.4. CR 201702763, NOS Insight EP Risk Significant Planning Standard Performance Upper Tier Cause Evaluation Needed Page l8of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 process, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant ora security event that involves probable life threatening risk to site pei-sonnel 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 applied to the Callaway ISfSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

Contaiiirnent 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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met. (ref. 4.1.9)

Emergency Action Level 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 One of a set of names or titles established by the US Nuclear Regulatoi-y Commission (NRC) for grouping off-normal events or conditions according to (1) potential or acttial 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) and General Emergency (GE).

EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: I Rem TEDE or 5 Rem CDI Thyroid. ActuaL or projected offsite exposures in excess of the EPA PAGs requires Callaway 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 circtiits, grotrnding, 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 teak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if Large quantities of smoke and heat are observed.

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

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

General Emergency Events are in process or have occurred which invoLve actuaL or imminent substantial core degradation or melting with potential fbr 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.

High Winds Winds in excess of 40 mph (18 rn/s) sustained, or 58 mph (26 m/s) gusting.

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

Hostile Action An act toward Callaway 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 he construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Callaway. 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 ttse of protective equipment, such as SCBAs, that is not routinely employed).

Independent Spent Fuel Storage tnstatlation (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 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.

Page 20 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Loss of Safety function (LOSF) A LOSE exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of the Safety Function Detennination Program, a LOSF may exist when a support system is mOPERABLE, AND:

a) A required system redundant to the system supported by the INOPERABLE stipport system is also INOPERABLE; OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR c) A required system redundant to the stippoll system for the supported systems (a) and (b) above is also INOPERABLE.

Maintain Take appropriate action to hold the value of an identified parameter within specified limits.

Owner Controlled Area (OCA) The fenced area contiguous to the Protected Area, designated by ArnerenUE (Callaway Plant) to be controlled for security purposes (ref 4.1.6).

Projectile An object directed toward a Nuctear Power Plant that couLd cause concern for its continued operability, reliability, or personnel safety.

Protected Area (P4) An area encompassed by physical barriers and to whicl access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88 100 Property-Site Layout, Owner Controlled Area and Suiioundin Area (ref. 4.1.7).

RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its nonnal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle darns). The RCS is capable of being placed in an intact condition by Operator Action, i.e.,

pressurized to support natural circulation cooling.

Reduced Inventory Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is lower than 3 feet beLow the Reactor Vessel flange (<64.0 in.) (ref. 4.1.13).

Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

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

Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

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 JOCFR5O.2):

Page 21 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 process or have occurred which involve actual or likely major failures of plant fitnctions needed for protection of the public or hostile actions that result in intentional damage or malicious acts; (I) 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 Guidelines exposure levels beyond the site boundary.

Site Boundary Exclusion Area Boundary is a synonymotts term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with IOCFR100 (ref. 4.1.12).

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

Unplanned A parameter change or an event that is not I) the resuLt of an intended evoLution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

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

t7alid 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 afected SAFETY SYSTEM train.

Page 22 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 5.2. Abbreviations/Acronyms Degrees Fahrenheit O

De%rees AC Alternating Current ATWS Anticipated Transient Without Scram Callaway Callaway Energy Center CDI Committed Dose Equivalent CFR Code of Federal Regulations CMT Containment CSFST Critical Safety function Status Tree DBA Design Basis Accident DBT Design Bases Threat DC ect Current EAL ... Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmentat Protection Agency EPIP Emergency Plan Implementing Procedure ERG Emergency Response Guideline 1SF Engineered Safety Feature 15W Essential Service Water 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 tndividual Plant Examination of External Events (Generic Letter 88-20)

Effective Neutron Multiplication Factor LCO Limiting Condition of Operation LER Licensee Event Report LOCA Loss of Coolant Accident LWR Light Water Reactor MPC Maximum Permissible Concentration/Multi-Purpose Canister mR, rnRern, mrem, mREM milli-Roentgen Equivalent Man MSL Main Steam Line MW Megawatt NIl Nuclear Energy Institute Page 23 of 239 INFORMATION USE

EIP-ZZ-00I0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 Offsite Dose Calculation Manual ORO Offsite Response Organization OTO Off-Normal Operating Procedure PA Protected Area PAG Protective Action Guideline PRA/PSA Probabilistic Risk Assessment / ProbabiListic Safety Assessment PWR Pressurized Water Reactor PSIG Pounds per Sqtiare Inch Gauge R Roentgen RCC Rectnr Control Console RCS 1rtnr Coolant System Rem, rem, REM Roentgen Equivalent Man RETS P Effluent Technical Specifications RPS Reictnr Protection System R(P)V Reactor (Pressure) Vessel RVLIS Reactor Vessel Level Indicating System SAR Analysis Report SBO.,,. Station Blackout SCBA Self-Contained Breathing Apparatus Steam Generator Safety Injection SPDS Safety Parameter Display System SRO Senior Reactor Operator 5Sf Safe Shutdown facility TEDE Total Effective Dose Equivalent TOAF Top of Active fuel TSC Technical Support Center WOG Westinghouse Owners Group Page 24 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 6.0 CALLAWAY-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Callaway EAL within the NET 99-01 IC/EAL identification scheme. further information regarding the development of the Callaway EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

Callaway NEI 99-0 1 Rev. 6 Callawa NE! 99-01 Rev. 6 Callaway NEI 99-01 Rev. 6 EAL RU1.1 IC AU1 Example 1,2 EAL CU5.1 IC CU5

] Example 1,2,3 EAL HA5.1 HA5 Example 1

RU1.2 AU1 3 CA1.1 CAl 1 HA6.1 HA6 1 RU2.1 AU2 1 CA1.2 CAl 2 HA7.1 HA7 1 RA1.1 AM 1 CA2.1 CA2 1 HS1.1 HS1 1 RA1.2 AA1 2 CA3.1 CA3 1,2 HS6.1 HS6 1 RA1.3 AA1 3 CA6.1 CA6 1 HS7,1 HS7 1 RA1.4 AM 4 CS1.1 CSJ 1 HG7.1 HG7 1 RA2.1 AA2 1 CS1.2 CS1 2 SU1.1 SU1 1 RA2.2 AA2 2 CS1.3 CS1 3 SU3.1 SU2 1 RA2.3 AA2 3 CG1.1 CG1 1 SU4.1 SU3 2 RA3.1 AA3 1 CG1.2 CG1 2 SU5.1 SU4 1,2,3 RA3.2 AA3 2 FA1.1 FA1 1 SU6.1 SU5 1 RS1.1 AS1 1 FS1.1 FS1 1 SU6.2 SU5 2 RS1.2 AS1 2 FG1.1 FG1 1 SU7.1 SU6 1,2,3 RS1.3 AS1 3 HU1.1 HU1 1,23 SU8.1 SU7 1, 2 RS2.1 AS2 1 HU2.1 HU2 1 SA1.1 SAl 1 RG1.1 AG1 1 HU3.1 HU3 1 SA3.1 SA2 1 RG1.2 AG1 2 HU3.2 HU3 2 SA6.1 SA5 1 RG1.3 AG1 3 HU3.3 HU3 3 SA9.1 SA9 1 RG2.1 AG2 1 HU3.4 HU3 4 SS1.1 SS1 1 CU1.1 CU1 1 HU4.1 HU4 1 SS2.1 SS8 1 CU1.2 CU1 2 HU4.2 HU4 2 SS6.1 SS5 1 CU2.1 CU2 1 HU4.3 HU4 3 SG1.1 SG1 1 CU3.1 CU3 1 HU4.4 HU4 4 SG1.2 SG8 1 CU3.2 CU3 2 HU7.1 HU7 1 EU1.1 E-HU1 1 CU4.1 CU4 1 HAJ.1 HAl 1,2 Page 25 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASIS DOCUMENT 7.0 ATTACHMENTS

7. 1. Attachment 1, Emergency Action Level Technical Bases 7.2. Attachment 2, Fission Product Barrier Loss / Potential Loss Matrix and Bases 8.0

SUMMARY

OF CHANGES Section or Step Page(s) Description uniber Changed the definition of Visible Damage to:

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.

to help clarify the intent of visible damage.

Added the following Definition:

Definitions Loss of Safety Function (LOSE) A LOSE exists when, assuming no concurrent CA6. 1 Def. single failure, a safety function assumed in the accident analysis cannot be 11= performed. For the purpose of the Safety Function Determination Program, a SA9,l Def I $7 LOSE may exist when a support system is INOPERABLE, AND:

A required system redundant to the system supported by the INOPERABLE support system is also INOPERABLE; OR

A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR

A required system redundant to the support system for the supported systems in bullets 1 and 2 above is also INOPERABLE.

25 6.0 Deleted FIG 1.1 from the Callaway to NE! cross reference chart.

69 EU 1.1 Basis Removed reference to FIG! from last sentence of basis.

Page 26 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Section or Step Page(s) Description Number Reworded the EAL to better address the intent. The new wording reads:

The occurrence of any Table C-6 (S-5) hazardous event AND Event damage has caused indication of degraded performance on train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating 111 CA6.I MODE.

& &

Event damage has resulted in VISIBLE DAMAGE to a second train of a 186 SA9. 1 SAFETY SYSTEM needed tor the current operating MODE.

(Notes 71, 72,1 And along with the revised EAL text, the following notes were also added:

Note 11: If the hazardous event only results in VISIBLE DAMAGE, with no indication of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Note 12: This EAL is applicable when a Table 0-6 (S-5), Hazardous Events, causes a LOSS OF SAFETY FUNCTION on a SAFETY SYSTEM required_for the current_operating_MODE.

Removed the existing basis text following the bulleted steps and replaced with the foIl owing:

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 othe:

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 ssues. Note that this second SAFETY SYSTEM train is from the same SAFETY 113 CA6.l Basis SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via Cs CS1 or RS1.

This was done to help clarify when the EAL should be called.

115 Ht. 1.1 BasisfRemoved reference to HGI from classifying HOSTILE ACTIONS..

120 HSI.1 Basis Removed reference to HGI and added all other General Emergency lOs.

Old HG 1.1 Deleted HG 1 1 and associated reference material from bases.

Page 27 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Section or Step Page(s) Description Number Removed the existing basis text follotving the paragraph about a single tacilted generator, and replaced with the fotlowing:

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 othe 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 188 SA9.1 Basis SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via ICs FS1 or RS1 This was done to help clarify when the EAL should be called.

Page 28 of 239 INFORMATION USE

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EIP-ZZ-OOlO1 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT This page intentionally left blank Page 30 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category R Abnornial Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable in All, Hot or Cold plant conditions.)

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 harriers 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 ola 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 he 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. Radiolothcal 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 ofa 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 precltide access to areas requiring continuous occupancy also warrant emergency classification.

Page 31 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategoty: I Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU 1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor> column UE for 60 mm.

(Notes I. 2. 3)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

tVote 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 GT-RE-21B 659E+7 pCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-1 11/112/

ASD Monitors (NBIC/D) 12 mR/hr 1.2 mR/hr 0

113/114 G

TD AFW Steam FC-RE-385 163 mRlhr 16.3 mR/hr 1.6 mR/hr Discharge Radwaste Bldg Vent GH-RE-1OB ---- ---- ---- 2 X Hi-Hi alarm Liquid Radwaste HB-RE-18 -- -- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Basis:

The column UI gaseous and Liquid release values in TabLe R-1 represent two times the appropriate ODCM reLease rate limits associated with the specified monitors (ref. 1, 2, 3).

Page 32 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The effluent monitor Hi-Hi alarm setpoints correspond to the Hi-Hi alarm (red) setpoint as displayed on RM-l1.

The Hi-Hi alarm value setpoints are available by examining channel 9 on the RM-23.

The RM- 11 Channel Number 2 13 is utilized for the Unit Vent (GT-RE-2 18) reading for Table R- 1. This channel is read out in iCi/sec while all others are read out in liCi/ml.

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 pennit 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, atid to control and monitor intentional releases. The occutrence 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 Ibr 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 ftilly 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, lithe effluent flow past an effluent tnonitor 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.

Callaway Basis Reference(s):

1. APA-ZZ-0 1003, Catlaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. FSAR Section 16.11.1.3, Radioactive Effluent Monitoring Instrumentation LCO

3. EPCI 1402, EAL TabLe R-1 Calculations

4. NIl 99-01, AU1 Page 33 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnotmal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL:

RU 1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for> 60 mm.

(Notes 1, 2)

Notc it 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 tinknown, assume that the release duration has exceeded the specified time limit.

MODE Applicability:

All Definition(s):

N one 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.

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 liqitids into storm drains, heat exchanger leakage in river water systems, etc.).

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.

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

Page 34 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases All water ninoff from the plant eventually flows into Logan Creek and then to the Missouri River. If radioactive liquid flows offsite, begin hourly grab samples at the Portland River Sample Location and analyze for tritiurn and gamma spectmm. Send results to the Dose Assessment Tecimician or Dose Assessment Coordinator for evaluation.

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

Callawav Basis Reference(s):

1. APA-ZZ-0 1003, Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NE199-01,AUI Page 35 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEIJ TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 1 Radiological Effluent Initiating Cotidition: Release of gaseous or liquid radioactivity resulting in offsite close greater than 10 nuem TEDE or 50 mrern thyroid CDE EAL:

RAI.1 Alert Reading on any Table R- I effluent radiation monitor> column ALERT for 15 mm.

(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 he 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 tiow 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.

Notc 4: The pre-calctilated effluent monitor values presented in EALs RAil, RS1.l and RGI.l should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-J Effluent Monitor Classification Thresholds Release Point Monitor GE

[ SAE Alert UE

]

Unit Vent GT-RE-21B 6.59E+7 pCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-l 11/112/

ASD Monitors (NB/CID) 12 mR/hr 1.2 mR/hr 0

113/114 C) 0 ID AFW Steam FC-RE-385 163 mRihr 16.3 mR/hr 1.6 mR/hr Discharge Radwaste Bldg Vent GH-RE-1 08 ---- ---- ---- 2 X Hi-Hi alarm Liquid Radwaste HB-RE-18 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 36 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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:

10 mRem TIDE

50 mRem CDI Thyroid 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.

To clarify Note 4, if a threshold value is met in Table R-l for a classification, there is a 15 minute time limit to make the cLassification (RA1 .1). If Dose Assessment (LrRI!RASCAL) is available it should he used instead (RA1.2) since it is more accurate than the values in Table R-1. However the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-1 value being exceeded OR the classification should be made using Table R-1 (RAI. I).

The RM-l I Channel Number 213 is utilized for the Unit Vent (GT-RE-21B) reading for Table R-l. This channel is read out in tCi!sec while all others are read out in jiCi/mi.

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 CDI Thyroid) (ref. 1).

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 tin-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 hilly addresses the spectrum of possible accident events and conditions.

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

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

Callawav Basis Reference(s):

1. EPCI 1402, EAL Table R-l Calculations

2. NET 99-01, AM Page 37 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Ernergeticy Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseotis 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 rnrern thyroid CDI at or beyond the SITE BOUNDARY.

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymotts term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

Dose assessments are performed by computer-based method (ref. 1, 2).

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input source. This value is in mrem TEDE or thyroid CDL This IC addresses a release of gaseous or liqtiid 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 close is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrern 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 effitient 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 RSI.

Page 38 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technica] Bases Callaway Basis Reference(s):

1. EIP-ZZ-01211, Accident Dose Assessment

2. EPCI 1402, EAL Table R-1 Calculations

3. NE199-0l,AAI Page 39 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad LeveLs / Rad Effluent Subcategory: I 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 rnrem thyroid CDE at or beyond the SITE BOUNDARY for 60 mm.

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 he 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 Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the Land suirounding the Ptaut to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCFRIOO.

Basis:

Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. I).

This IC is based on liquid sample analysis by the Count Room.

This IC addresses a release of gaseous or tiquid 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 incltided 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.

Page 40 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Techtiical 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.

All water runoff from the plant eventually flows into Logan Creek and then to the Missouri River. If radioactive liquid flows offsite, begin hourly grab samples at the Portland River Sample Location and analyze for tritium and gamma spectrum. Send results to the Dose Assessment Technician or Dose Assessment Coordinator for evaluation.

Escalation of the emergency classification level would he via IC RS 1.

Callaway Basis Reference(s):

1. APA-ZZ-0 1003, Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NFl 99-01, AM Page 41 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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:

RAI.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 fork 60 mm.

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

(Notes 1. 2)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely he exceeded, Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specifled time limit.

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land sutTounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCfR100.

Basis:

EIP-ZZ-0021 1, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. I).

This IC is based solely on field monitoring team restdts without performing calculations using the Unified RASCAL Interface software (URI).

The closed window valite is in mRihr. The analysis of field survey samples is in mrern thyroid CDE for 60 minutes.

This IC addresses a release of gaseous or Liquid radioactivity that restilts 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).

Page 42 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Radiological effluent EALs are also included to provide a basis for classifying events anti 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 ftilly addresses the spectrum of possible accident events and conditions.

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

Classification based on effluent monitor readings assumes that a release path to the enviromnent 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 RSI.

Callaway Basis Reference(s):

1. EIP-ZZ-00211. Field Monitoring

2. NIl 99-01, AA1 Page 43 of 239 INFORMATION USE

EIP-ZZ-OOlOl ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 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 mm.

(Notes I, 1. 3, 4)

Note I: 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 stall 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-calcitlated effluent monitor values presented in EALs RAT .1. RSI.l antI RG1.l should be used for emergency classification assessments tintil the results from a dose assessment using actual meteorology are avai Table.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE J Alert UE

]

Unit Vent GT-RE-21B 6.59E+7 pCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-1 11/112/

ASD Monitors (NBIC/D) 12 mRhr 1.2 mRJhr --

0 113/114 G

ID AAN Steam 1 6.3 mR/hr 1.6 mR/hr 0 FC-RE-385 163 mR/hr Discharge Radwaste Bldg Vent GH-RE-1OB -- -- 2 X Hi-Hi alarm Liquid Radwaste HB-RE-18 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 44 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Ret. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

This EAL address gaseous radioactivity releases, that for whatever reason. cause effluent radiation monitor readings corresponding to stte boundary doses that exceed either:

100 mRem TEDE

500 mRem CDE Thyroid Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flotv 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.

To clarify Note 4, if a threshold vaitte is met in Table R-1 for a classification, there is a 15 minute time limit to make the classification (RS 1 .1). If Dose Assessment (URI/RASCAL) is available it should be used instead (RS1.2) since it is more accurate than the values in Table R-l. However the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-1 value being exceeded OR the classification should be made using Table R-l (RSI.l).

The RM- 11 Channel Number 213 is utilized for the Unit Vent (GT-RE-2 I B) reading for Table R- 1. This channel is read out in liCi/sec while all others are read out in uCi/mI.

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).

This IC addresses a release of gaseotis 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 tvith 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 ftilly addresses the spectnim of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 rnrem 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 RGI.

Callaway Basis Reference(s):

1. EPCI 1402. EAL Table R-l Calcutations

2. NE199-0I,AS1 Page 45 of 239 INFORNIATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUfVIENT Attachment 1 - Emergency Action Level Technical Bases Category:R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 nwem TEDE or 500 mrem thyroid CDE EAL:

RSI.2 Site Area Emergency Dose assessment using actual meteorology indicates doses> 100 imem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY.

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the Land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 10CFR 100.

Basis:

Dose assessments are performed by computer-based method (ref 1, 2)

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input sotirce. This value is in mrem TEDE or thyroid CDE.

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 tin-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 rnrem thyroid CDE was established in consideration othe 1:5 ratio of the EPA PAG for TEDE and thyroid CDI.

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.

Page 46 of 239 INFORMATION USE

E[P-ZZ-00l01 ADDENDUM 2 Rev. 016 ENIERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. EIP-ZZ-0 1211, Accident Dose Assessment

2. EPCI 1402. EAL Table R-1 Calculations

3. N1199-01,AS1 Page 47 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I -- RadioLogical Effluent Initiating Condition: Release of gaseous tadioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 rnrern 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 mitt.

Analyses of field survey samples indicate thyroid CDE> 500 mrem for 60 mm. of inhalation.

(Notes I. 2)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

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

MODE Applicability:

Alt Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 tet) from the midpoint of the Unit 1 Reactoi- Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCFRIOO.

Basis:

EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. I).

This IC is based solely on field monitoring team resuLts without performing calculations tising the Unified RASCAL Interface software (URI).

The closed window value is in mRihr. The analysis of field survey samples is in mrern thyroid CDE for 60 minutes.

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

Page 48 of 239 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 ftilly addresses the spectrum of possible accident events and conditions.

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

Escalation of the emergency classification level would be via IC RG I.

Callaway Basis Reference(s):

1. EIP-ZZ-002 11, field Monitoring

2. NEI 99-01, ASL Page49 of 239 INFOR1/IATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Siibcategory: 1 Radiological Effluent Initiating Condition: ReLease of gaseous radioactivity resulting in offsite dose greater than 1,000 mrern TEDE or 5,000 mrern thyroid CDE EAL:

RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor> column GE for 15 mm.

(Notes I, 2. 3. 4)

Note I: The Emergency Coordinator should declare the event proinptly 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 RAil, RSI.1 and RG1.l should be used for emergency classilication 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 GT-RE-21B 6.59E+7 pCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-iii/112/

ASD Monitors (NB/C/D) 12 mR/hr 1.2 mRlhr TDASteam FC-RE-385 163mR/hr 16.3mR/hr 1.6mR/hr Discharge Radwaste Bldg Vent GH-RE-1OB ---- -- ---- 2 X Hi-Hi alarm Liquid Radwaste HB-RE-18 ---- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 50 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 tnRem TEDE

50t)0 rnRem CDE Thyroid 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.

To clarify Note 4, if a threshold value is met in Table R-l for a classification, there is a 15 minute time limit to make the classification (RGI.l). If Dose Assessment (IJRLRASCAL) is available it should be used instead (RG1.2) since it is more accurate than the values in Table R-l. However the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-l value being exceeded OR the classification should be made using Table R-l (RG1.l).

The RM-1 I Channel Number 213 is utilized for the Unit Vent (GT-RE-21 B) reading for Table R-1. This channel is read out in tCisec while all others are read out in tiCi/mI.

The column GE gaseous effluent release values in Table R-l correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TIDE or CDE Thyroid) (ref. 1).

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 Gtiicles (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 ftdly addresses the spectrum of possible accident events and conditions.

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

Callaway Basis Reference(s):

1. EPCI 1402, EAL Table R-I Calculations

2. NET 99-01, AGI Page 51 of 239 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / Rad Effitient Snbcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrern TEDE or 5,000 rurem thyroid CDI EAL:

RG1.2 General Emergency Dose assessment using actual meteorology indicates doses> 1,000 mrern TEDE or 5,000 mrern thyroid CDE at or beyond the SITE BOUNDARY.

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land sunotLnding the Plant to a radius of I ,200 meters (3,937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 1OCfR 100.

Basis:

Dose assessments are performed by computer-based method (ref. 1, 2)

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input source. This value is in rnrem TEDE or thyroid CDE.

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 I ,000 mrem while the 5,000 mrem thyroid CDI was established in consideration of the 1:5 ratio of the EPA FAG 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.

Page 52 of 239 INFORMATION USE

EIP-ZZ-OO 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. EIP-ZZ-t)121 1, Accident Dose Assessment

2. EPCI 1402, EAL Table R-1 Calculations

3. NFl 99-01, AG1 Page 53 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R - Abnormal Ract Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrern TIDE or 5,000 mrem thyroid CDI EAL:

RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates> I .000 mR/hr expected to continue for? 60 mm.

Analyses of field survey sampLes indicate thyroid CDE > 5,000 mrem for 60 mm. of inhalation.

(Notes 1, 2)

Note I: 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 dciration has exceeded the specified time limit.

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a racLius of 1,200 meters (3,937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCfRIOO.

Basis:

EIP-ZZ-0021 1, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC is based solely on field monitoring team results without performing calculations using the Unified RASCAL Interface software (URI).

The closed window value is in rnRlw. The analysis of field survey samples is in mrem thyroid CDI for 60 minutes.

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 (PAG5). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent FALs 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.

Page 54 of 239 INFORIIATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at the EPA PAG of 1,000 mrern while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDL Callawav Basis Reference(s):

1. EIP-ZZ-002 11, Field Monitoriiw

2. NIl 99-01, AGI Page 55 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 (EQ LI-0039A, EC LI-0039B, local observation of SF P level).

AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2 radiation monitors.

Table R-2 Fuel Building & Containment Area Radiation Monitors Fuel Building: Containment:

SD-RE-34, Cask Handle Area Radiation

SD RE 40, Personnel Access Hatch Area

SD-RE-35, New Fuel Storage Area Radiation

SD RE 41, Manipulator Crane Radiation Monitor

SD-RE-36, New Fuel Storage Area Radiation

SD RE 42, Containment Building Radiation

SD-RE-37, Fuel Pool Bridge Crane Radiation

GT RE 59 Containment High Area Radiation Monitor

SD-RE-38, Spent Fuel Pool Area Radiation

GT RE 60 Containment High Area Radiation Monitor 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 PA THWA Y The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

Basis:

The low water level alarm in this EAL refers to the Spent Fuel Pool (SfP) low level alarm (Annunciator 76D, SF? LEV HI LO) (ref. 1). During the fuel transfer phase of refueling operations, the fuel transfer canal is normally in communication with the spent fuel pool and the refueling pool in the Containment is in communication with the fuel transfer canal when the fuel transfer tube is open. A lowering in water level in the SF?, fuel transfer canal or refueling pool is therefore sensed by the SFP low level alarm. Neither the refueling pool nor the fuel transfer canal is equipped with a low level alarm (ref. 1).

The SFP level is remotely monitored by level indicator EC LI-0039A. The level switch initiates high and low level annunciators Page 56 of 239 INFORMATION USE

EIP-ZZ-OOlOl ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Technical Specification 3.7.15 (ref. 2) requires at least 23 ft. of water above the Spent Fuel Pool storage racks. Teclrnical Specification 3.9.7 (ref. 3) requires at least 23 ft. of water above the Reactor Vessel flange in the refueling pool. During refueling, this maintains sufficient water level in the fuel transfer canal, refueling pooi. and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident.

The Table R-2 radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventoly (ref. 1). 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.

This IC addresses a decrease in water level above irradiated fuel sufficient to caitse 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 detetinined 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 unplatrnecl Loss of water level.

A drop in water level above iftadiated 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.

Callaway Basis Reference(s):

I. OTO-EC-0000l, Loss of Spent Fuel Pool/Refuel Pool Level

2. Technical Specification 3.7.15, Fuel Storage Pool Water Level

3. Technical Specification 3.9.7, Refueling Pool Water Level

4. NIl 99-01, AU2 Page 57 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 PA THWA Y The reactor reftieling cavity, spent fuel pooi 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 pooi. 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.l in that the loss of level, in the affected portion of the REFUELING PATI-IWAY, 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 ilTadiated fuel within the reactor vessel may require classification in accordance Recognition Category C during the Cold Shutdown and Reftteling MODES.

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

Callaway Basis Reference(s):

1. OTO-EC-00001, Loss of Spent fuel Pool/Refuel Pool Level

2. NIl 99-01, AA2 Page 58 of 239 INFORIIATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad LeveLs / Rad Effluent Subcategory: 2 ItTacliated fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated the!

EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by any of the following:

Hi-Hi Alarm on fuel Building exhaust monitors (GG-RE-27 or 28).

Manipulator crane radiation monitor (SD-RE-41) >100 mRhr.

Fuel Pool Bridge Crane OR Spent fuel Pool Area radiation monitor (SD-RE-37 or

38) > 30 mR/hr.

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 ftiel (tel. 1, 2).

The bases for the SF? ventilation radiation Hi-Hi alarm and the SFP and containment area radiation readings are a spent fuel handling accident (ref. 2, 3). In the Fuel Handling Building, a fuel assembly could be dropped in the fuel transfer canal or in the SFP. Shotild a fuel assembly be dropped in the fuel transfer canal or in the SFP and release radioactivity above a prescribed level, the fuel handling building ventilation monitors sound an alarm, alerting personnel to the problem (ref. 1,2, 3.4).

This IC addresses events that have caused imminent or actual damage to an itTadiated 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 lUll. Cask is sealed when welding is complete.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

Escalation of the emergency classification level would be via IC RS1.

Page 59 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. OTO-EC-0000l, Loss of Spent Fuel Pool/Reftiel Pool Level

2. OTO-KE-0000I, Fuel Handling Accident

3. (ale. EPCI 98-0 1, Emergency Action Level Bases

4. CaIc. HPCI 05-02, Gaseous and Liquid Radiation Monitor Setpoints

5. NEI 99-01, AA2 Page 60 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 2 Irradiated Fttel Event Initiating Conditioii: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 120 as indicated on EC-L1-0059A or EC-LI-0060A.

MODE Applicability:

All Definition(s):

None Basis:

Post-fukushima order EA- 12-05 1 (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 ftiel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

for Callaway Plant SFP Level 2 is plant elevation 2031 ft. 1 .25 in. (9 ft. 11 in. above the top of the spent fuel racks) as indicated by 120 on EC-L1-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-L1-0060A in the Auxiliary Building haLlway 2026.

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 acttial 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 pooi. 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 RS2.

Callaway Basis Reference(s):

1. NRC EA-12-51, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. SFPIS Mod Overview for EP 3 NEI 99-01, AA2 Page 61 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / RacI Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pooi level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pooi level to 12 as indicated on EC-LI-0059A or EC-LI-0060A.

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 I), SFP level 10 ft. above the top of the fuel racks (Level 2) and SF P level at the top of the fttel racks (Level 3).

For Callaway Plant SFP Level 3 has been set at a plant elevation 2022 ft. 1.25 in. (--1 1111. above the top of the spent fuel racks) as indicated by 12 on EC-Ll-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-Ll-0060A in the Auxiliary Building hallway 2026.

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMTNENT 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 RGI or RG2.

Callaway Basis Reference(s):

1. NRC EA- 12-51, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. SFPIS Mod Overview for EP

3. NE199-0l,A52 Page 62 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Ahnonnal 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 pooi level cannot be restored to at least 12 as indicated on EC-LI-0059A or EC-LI-0060A for> 60 mm.

(Note 1) iVote 1: The Emergency Coordinator should declare the eent promptly upon determining that time limit has been exceeded, or will likely he exceeded.

MODE Applicability:

All Definition(s):

None Basis:

Post-f ukushima order EA-l2-051 (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).

for Callaway Plant SF? Level 3 has been set at a plant elevation 2022 ft. 1 .25 in. (1 1 in. above the top of the spent fuel racks) as indicated by 12 on EC-LI-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-L1-0060A in the Auxiliary Building hallway 2026.

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged tmcovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

tt is recognized that this IC would likely not be met until well after another General Emergency IC was niet however, it is included to provide classification diversity.

Callawav Basis Reference(s):

1. NRC EA-12-51, Issuance of Order to Modify Licenses with Regard to Reliable Spent Ftiel Pool Instmrnentation

2. SFPIS Mod Overview for EP

3. NE199-0l,AG2 Page 63 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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.l Alert Dose rates> 15 mR/hr in EITHER of the following areas:

Control Room (SD-RE-33).

Central Alarm Station (by survey).

MODE Applicability:

All Definition(s):

None Basis:

Areas that meet this threshoLd include the Control Room and the Central Alarm Station (CAS). SD-RE-33 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 (ref. 1).

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to precltide 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 actttat 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.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

Callaway Basis Reference(s):

I. FSAR Section 12.3, Table 12.3-2, Area Radiation Monitors

2. NET 99-0 1, AA3 Page 64 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels I 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 EITHER of the following: (iVote 5)

North Electrical Penetration Room. (Room 1410)

South Electrical Penetration Room. (Room 1409)

Vote 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.

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).

u\rPLLVNED A parameter change or an event that is not I) 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 only rooms/areas external to the Control Room that require access to perform field actions consistent with the above criteria for Callaway are the North and South Electrical Penetration Rooms when in MODE 4 to support isolating SI accumulators and placing RHR in service for RCS cootdown to Cold Shutdown (ref.

1, 2. 3). The equipment required is:

For SI Accumulators:

NGOIBGF3, FDRBKRTOEPHVSXOSASI ACCAOLT ISO. (Room 1410)

NGO2BGF3. FDR BKR TO EPHVSSOSB St ACC B OUT ISO. (Room 1409)

NGO1BGF2. FOR BKR TO EPHVSSO$C St ACC C OUT ISO. (Room 1410]

NGO2BHF2, FUR BKR TO EPHVSO8D St ACC D OCT ISO. (Room 1409]

for A RHR:

NGO2BCF2, FDR BKR TO BBPVS7O2A RCS LOOP 1 HOT LEG TO RHR PMPS ISO. (Room 1409)

NGO1BEF2, FUR BKR TO EJHV87t)IA A RHR PMP SUCT FROM RCS HOT LEG 1ISO. (Room 1410) for B RHR:

NGO2BBF3, FDR BKR TO BBPV87O2B RCS LOOP 4 HOT LEG TO RHR PMPS ISO, (Room 1409)

NGO1BDF3, FUR BKRTO EJHVX7O1B B RHR PMP SUCT FROM RCS HOT LEG 4ISO, (Room 1410)

Page 65 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 acti.tal 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.

Alert declaration is warranted if entry into the affected rootnlarea 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 tipon whether entry is actually necessary at the time of the increased radiation levels.

Access shotild 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 nornial administrative limits).

An emergency declaration is not warranted if any of the following conditions apply:

The plant is NOT in MODE 4.

The increased radiation levels are a restdt 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 totinds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

Callaway Basis Reference(s):

1. OTG-ZZ-00006 Addendum 06, Securing Safety Injection Accumulators 2 OTN-EJ-00001 Addendum 3, Placing A RHR Train In Service for RCS Cooldown

3. OTN-EJ-0000l Addendum 4, Placing B RHR Train In Service for RCS Cooldown

4. NEI 99-01, AA3 Page 66 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category E Independent Spent fuel Storage Installation CISFSI)

EAL Grotip: Any tEALs in this category are applicable to All, Hot, or Cold plant conditions.)

An independent spent fuel storage installation (ISFSI) is a facility that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated.

Page 67 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: E Independent Spent fuel Storage Installation (ISFSI)

Subcategory: I - Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EUI.1 Unusual Event Damage to a loaded cask CONFINEMENT BOuNDARY as indicated by an on-contact radiation reading > EITHER of the following:

60 rnrern/hr (gamma + neutron) on the top of the closure lid of the Overpack!VVM.

7,000 mrem/hr (gamma + neutron) on the side of the Transfer Cask.

MODE Applicability:

All Definition(s):

CONFINEMENTBOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the Callaway ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

OVERPACK For the HI-Storm UMAX, the term OVERPACK is synonyms with the term VVM.

TRANSFER cASK Containers designed to contain the MPC during and alter Loading otspent fuel assemblies, and prior to and during unloading and to transfer the MPC to or from the overpack/VVM VERTICAL VENTILA TED MODULE (VIM,) A subtetTanean type overpack which receives and contains the sealed MPC for interim storage at the ISF$I. The VVM supports the MPC in a vertical orientation and provide gamma and neutron shielding and also provides air flow through cooling passages to promote heat transfer from the MPC to the environs.

Basis:

Confinement boundary is established at Callaway when the Multi-Ptirpose Canister welding is complete.

Overpacks/VVM casks receive and contain the sealed MPCs for interim storage in the ISFSI. They provide gamma and neutron shielding, and provide for ventilated air flow to promote heat transfer from the MPC to the environs. The term overpacklVVM does not include the transfer cask (ref. 1).

The values shown represents 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification 5.3.4 for radiation external to either a loaded MPC overpacklVVM or transfer cask (ref. 1).

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated ftiel 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 cask is sealed when the welding is complete.

Page 68 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The existence of damage is determined by radiological survey. The technical specification multiple of2 times. which is also used in Recognition Category R EQ RUY, 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 tate. 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 IC HS 1.

Callawav Basis Reference(s):

1. Certificate of Compliance No. 1040 Appendix A Technical Specifications for the HI-STORM UMAX Canister Storage System

2. NFl 99-0 1, E-HUI Page 69 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Cateorv C Cold Shutdown / Refueling System Malfunction EAL Grotip: Cold Conditions (RCS temperature 200°F); 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 occtirs 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 Sluitdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment cLosure, and fuel clad integrity for the applicable operating MODES (5 Cold Shutdown. 6 Reftteling, D Deftieled).

The events of this category pertain to the following subcategories:

I. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of emergency plant electrical Power can compromise plant safety system operability including decay heat removal and emel-gency core cooling systems which may he necessary to ensure fission product harrier integrity. This category includes loss of onsite and offsite power sources for

4. 16KV AC emergency buses.

3. RCS 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 etectrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission prodtict barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.

5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential persomiel within or external to the plant warrant emergency classification.

6. Hazardous Event Mfecting Safety Systems Certain hazardous nati.tral and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification.

Page 70 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CU 1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for 15 mm.

(Note I)

Note I: 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:

With the plant in Cold Shutdown, RCS water level is normally maintained above the pressurizer low level setpoint of l7°/ (ref. 1). However, if RCS level is being controlled below the presstlrizer tow 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 RCS that is the concern.

With the plant in Refueling MODE, RCS water level is normally maintained at or above the reactor vessel flange (Technical Specification 3.9.7 requires at least 23 ft. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters cruciaL to RCS draining operations (ref. 3).

This IC addresses the inability to restore and maintain water level to a required minimum level. This condition is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RCS 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.

Page 71 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases This EAL recognizes that the minimum required RCS 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 RCS level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer.

The 1 5-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 toss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.

Caltaway Basis Reference(s):

1. F R-I.2, Response to Low Pressurizer Level

2. OTN-BB-00002, Reactor Coolant System Draining

3. Technical Specification 3.9.7, Reftieling Pool Water Level

4. NEI 99-01, CU1 Page 72 of 239 INFORMATION USE

EIP-ZZ-001tJ1 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUNIENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malftinction Subcategory: 1 RCS Level Initiating Cotidition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CU1.2 Unusual Event RCS water level cannot be monitored AND EITHER

UNPLANNED increase in an Table C-I sump/tank level due to loss of RCS inventory.

Visual observation of UNISOLABLE RCS leakage.

Table C-I Sumps I Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

RCS INTACT The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seats or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

UNISOLABLE - An open or breached system line that cairnot be isolated, remotely or locally.

UNPLANNED A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

In Cold Shutdown MODE, the RCS will normally be intact and standard RCS level monitoring means are available.

In the Refuel MODE, the RCS is NOT intact and RPV level may be monitored by cli fferent means, including the abilth to monitor level visually.

Page 73 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases In this EAL, all water level indication is unavailable and the RCS inventory loss must be detected by indirect leakage indications. 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 RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even lithe source of the teakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

This IC addresses a loss of the ability to monitor RCS level concurrent with indications olcoolant leakage.

This condition is considered to be a potential degradation of the level of safety of the plant.

Refueling evoltitions that decrease RCS 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 level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-I). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage fiom the RCS.

Continued loss oURCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.

Callaway Basis Reference(s):

1. OTO-BB-00003, R0l4, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. NEI 99-01, CUI Page 74 of 239 INFORMATION USE

EIP-ZZ-t)0101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEt7EL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 1 RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.1 Alert Loss of RCS inventoiy as indicated by Reactor Vessel level < bottom of RCS hot leg ID (RVLIS Pumps Off< 73°/s).

MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

None Basis:

When Reactor Vessel water level towers to 2013.29 ft. (ref. 1), the inside diameter (ID) of the bottom of the RCS hot leg penetration is uncovered. The elevation of the bottom of the RCS hot leg penetration can be monitored only by RVLIS. (Note that this threshold is the loop penetration at the Reactor Vessel not the low point of the loop.) (ref. 3) When RVLIS is out of service, classification should be based on CAI.2 if RCS inventory cannot be monitored.

The RVLIS Pumps Off threshold has been determined as follows (ref. 1.2):

Elevation of bottom of Reactor Vessel (It) A 1987.150 Elevation of bottom ID of RCS hot leg penetration (ft) B 201 3.290 Hot leg penetration (above vessel bottom) C = B A (ft)

- 26.140 Height of vessel D (It) 41.245 RVLIS indication corresponding to the top of the core: H = 100 x C I D (%) 63.377 RVLIS overall channel accuracy: OCA 7.48% + (0.0104 x H) + 0.81%

OCA at H (%) 8.949 Bottom ID of RCS loop, including channel uncertainties: H + OCA (3/4) 72.327 Rounded upward to nearest 1% (RVLIS range is 0- 120% in 2% increments) 73 The threshold was chosen because level indication may be lost (RVLIS is normally inoperable in Refueling MODE (ref. 2)) and loss of suction to decay heat removal systems has occulTed. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier.

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 tevel of plant safety.

Page 75 of 239 INfORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases For this EAL, a towering of RCS water tevel below the specified level indicates that operator actions have not been successftit in restoring and maintaining RCS water tevet. 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 RCS 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 RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If RCS water level continues to lower, then escalation to Site Area Emergency would be via IC CSI.

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177 (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. NEI 99-01, CAl Page 76 of 239 INFORMATION USE

EIP-ZZ-OO 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 1 RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.2 Alert RCS water level cannot be monitored for? 15 mm. (Note l, AND EITHER

UNPLANNED increase in any Table C-l Sump / Tank level.

Visual observation of UNISOLABLE RCS leakage.

iVote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or tvill likely be exceeded.

Table C-i Sumps I Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump MODE Applicability:

5 - Cold Shutdown, 6 Refueling Definition(s):

RCS INTACT- The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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

UNPLAILVED A parameter change or an event that is not 1) the resuLt of an irnendecl evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

In Cold Shutdown MODE, the RCS will normally be intact and standard RCS level monitoring means are available.

In the Refuel MODE, the RCS is NOT intact and RPV level may be monitored by different means, incLuding the ability to monitor level visually.

Page 77 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases In this EAL all RCS water level indication would be tinavailable for greater than 15 minutes, and the RCS inventory loss must he detected by indirect leakage indications (Table C-I). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances.

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 RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

This IC addresses conditions that are prectirsors to a Loss of the ability to adeqtiately cool ilTadiated fitel (i.e.,

a precursor to a challenge to the fuel clad batTier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL, the inability to monitor RCS level may be catised 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 RCS.

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 RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI.

Callaway Basis Reference(s):

1. OTO-BB-00003-R014, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. NFl 99-01, CAl Page 78 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 1 RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS 1.1 Site Area Emergency With CONTAINMENT CLOSURE not established, RVLIS Pumps Off< 72%.

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 harrier to fission product release tinder shutdown conditions.

As appLied to Callaway, Containment Closure is established when the requirements of OSP-GT-00003, Containment Closure are met.

Basis:

When Reactor Vessel water tevel lowers to 2012.79 ft. (ref. I), water Level is six inches below the bottom of the RCS hot leg penetration. When Reactor Vessel water leveL drops significantLy below the bottom of the RCS hot leg penetration, all sources of RCS injection have failed or are incapable of making up for the inventory loss. Six inches below the bottom of the RCS hot leg penetration can be monitored only by RVLIS. Level monitoring instruments BB LI-53A/B and Computer Point BBLOO53BB cannot sense level changes in the Reactor Vessel below the RCS loop hot leg penetration. The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should be based on CS 1.3 if RCS inventory cannot be monitored.

The RVLIS Pumps Off threshold has been determined as follows (ref 1, 2):

Elevation of bottom of Reactor Vessel (if) A 1987.150 Elevation of bottom ID of RCS hot leg penetration ftt) B 2013.290 Six inches below hot leg penetration (above vessel bottom) C = B A 0.5 (if)

- - 25.640 Height of vessel D fft) 41.245 RVLIS indication corresponding to the top of the core: H = 100 xC ID (%) 62.165 RVLIS overall channel accuracy: OCA = 7.48% ÷ (0.0104 x H) + 0.81%

OCA at H (%) 8.937 Six inches below Bottom ID of RCS loop, including channel uncertainties: H + OCA (3/4) 71.102 Rounded upward to nearest 1% (RVLIS range isO 120% in 2% increments) 72 Page 79 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control. tnventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4).

This IC addresses a significant and proLonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor cooLant boiling and a further reduction in reactor vesseL level. If RCS level cannot be restored, fttet damage is probabLe.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE foLlowing a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs CS I l and CS 1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CGI or RG1.

Callaway Basis Refercncc(s):

I. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177, (387.1 -CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Containment Clostire

5. NE199-Ol,CSI Page 80 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS 1.2 Site Area Emergency With CONTAINMENT CLOSURE established. RVLIS Pumps Off< 65% (lop of Fuel).

MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

CONT/IINMENT CLOSURE The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated stntctures, systems, and components as a functional hauTier to fission product release under shutdown conditions.

As applied to Caltaway, Containment Closure is established when the requirements of OSP-GT-00003, Containment CLosure are met.

Basis:

When Reactor Vessel water level drops below RVLIS Ptirnps Off indication of 65% (2010.29 ft.), core uncovery is about to occur. The Plant Computer System Display caLled RefLel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should he based on CSI .3 if RCS inventory cannot be monitored.

The RVLIS Pumps Off threshold has been cletennined as follows (ref. 1, 2):

Elevation of bottom of Reactor Vessel (ft) A 1987.150 Elevation of top of fuel (ft) B 2010.290 Height of top of core (above vessel bottom)C = B-A (if) 23.140 Height of vessel D (ft) 41 .245 RVLIS indication corresponding to the top of the core: H = lOOx C/D(°Jo) 56.104 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81%

OCA at H (%) 8.873 Top of core, including channel uncertainties: H + OCA (%) 64.977 Rounded upward to nearest 1% fRVLIS range is 0- 120% in 2% increments) 65 Page 81 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative ola loss of inventory control. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this toss of water indicates that makeup systems have not been effective and may not be capable of preventing fttrthcr RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the fuel Clad barrier.

The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4).

This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant fttnctions 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 RCS level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs CS 1 .1 and CS 1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fissiOn prodtict release to the environment.

This EAL addresses concerns raised by Generic Letter 88-17, Loss o Decay 1-leat Removal; SECY 9 1-283, 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 9 1-06, Gtiidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CGI or RGI.

Caltaway Basis Reference(s):

I. OOA-BB-00003, Refuel Level Indications

2. Catcttlation No. BB-177, (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Containment Closure

5. NE199-01,CS1 Page 82 of 239 INFOR1VIATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 1 RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RCS water level cannot he monitored for> 30 mm. (Note 1)

AND Core uncovery is indicated by any of the following:

UNPLANNED increase in any Table C-i sump/tank level of sufficient magnitude to indicate core uncovery.

Manipulator crane radiation monitor SD-RE-41 > 10,000 mR/hr.

Erratic Sotirce Range Monitor indication.

Vote 1: The Emergency Coordinaior should declare the event promptly upon determminu that time limit has been exceeded, or will likely he exceeded.

Table C-i Sumps I Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump MODE Applicability:

5 - Cold Shutctown, 6 Refueling Definition(s):

RCS INTACT The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

UVPLAIV7vTfD A parameter change or an event that is not 1) the restdt 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 tinknown.

Basis:

In Cold Shutdown MODE, the RES will normally be intact and standard RCS level monitoring means are available.

Page 83 of 239 INFORMATION USE

E[P-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases In the Refueling MODE, the RES is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RES water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-I). SurveiLlance procedtires provide instructions for calculating primary system teak rate by manual or computer-based water inventory balances.

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 RCS leakage. If the make-up rate to the RES unexplainably rises above the pre-established rate, a loss of RES inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RES that cannot he isolated could also be indicative of a loss of RES inventory (ref. 1,2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RES draining operations (ref. 3).

The Reactor Vessel inventory loss may be detected by the manipulator crane radiation monitor or erratic Source Range Monitor indication. As water level in the Reactor Vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in up-scaled manipulator crane radiation monitor (SD-RE-41) indication (ref. 4, 5, 6).

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 (ref. 7, 8).

This IC addresses a significant and prolonged loss of RES inventory control and makeup capability leading to IMMINENT ftmel damage. The lost inventory may be due to a RES 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 RES 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 colTelate 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 RES 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 RES.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SEEY 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 NUMARE 9 1-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC EGl or RGI Page 84 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Techiiical Bases Callaway Basis Reference(s):

1. OTO-BB-00003-R014. Excess RCS Leakage

2. OSP-BB-00009, RCS [nventory Balance

3. OTN-BB-00002. Reactor Coolant System Draining

4. FSAR. Section 12.3.3.4

5. FSAR, Table 12.3-2

6. CaIc. No. HPCI -0701, SD-RE-41 Response to Core Uncovery in Refueling MODE

7. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects, pgs. 2-18, 2-19

8. Nuclear Safety Analysis Center (NSAC), 1980, Analysis of Three Mile Island Unit 2 Accident, N SAC-I

9. NEI 99-01. CS1 Page 85 of 239 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventory affecting fttel clad integrity with containment challenged EAL:

CG1.1 General Emergency RVLIS Pumps Off< 65% (Top of Fuel) fork 30 mm. (Now)

AND Any Containment Challenge indication, Table C-2.

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likeLy be exceeded.

Note 6: IfCONTALNMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration ofa General Emergency is not required.

Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)

Containment hydrogen concentration 4%

Unplanned rise in Containment pressure 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 stntctures, systems, and components as a functional harrier to fission product release under shutdown conditions.

As applied to Callaway, Containment Closure is established when the requirements oOSP-GT-00003 Containment Closure are met.

Basis:

When Reactor Vessel water level drops below RVLIS Pumps Off indication of 65% (2010.29 ft.), core uncovery is about to occur. The Plant Computer System Display called Refuel Levet Indications (turn on code Rh) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should be based on CG1.2 if RCS inventory cannot be monitored.

Page 86 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TEChNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases The RVL1S Pumps Off threshold has been determined as follows (ref. 1, 2):

Elevation of bottom of Reactor Vessel (if) A 1987.150 Elevation of top of fidel (if) B 2010.290 Height oftop of core (above vessel bottom)C = B -A (ft) 23.140 Height of vessel D (ft) 41 .245 RVLIS indication corresponding to the top of the core: H = 100 x C / D (%) 56.104 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81%

OCA at H (%) 8.873 Top of core, including channel uncertainties: H + OCA (%) 64.977 Rounded upward to nearest 1% (RVLIS range is 0 - 120% in 2% increments) 65 Three conditions are associated with a challenge to Containment integrity:

I. CONTAINMENT CLOSURE not established The status of Contaimnent closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4). If contaimnent closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

2. Containment hydrogen 4% The 4°/b hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. Callaway is equipped with a Hydrogen Control System (HCS) which serves to limit or reduce combustible gas concentrations in the Containment. The HCS is an engineered safety feature with redundant hydrogen recombiners, hydrogen mixing system, hydrogen monitoring subsystem, and a backup hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref. 5). Two Containment hydrogen monitors (GS Al-b and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and Emergency Response Facilities Information System (ERFIS) (ref. 6, 7). The hydrogen monitors require a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 8, 9).

3. UNPLANNED rise in Containment pressure An unplanned pressure rise in containment while in CoLd Shutdown or Refueling MODES can threaten Containment Closure capability and thus Containment potentially cannot be relied upon as a barrier to fission product release (ref. 4).

Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control with a challenge to the Containment. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level inventory within 30 minutes after reaching this condition in combination with a Containment challenge infers a failure of the RCS barrier, Loss of the Fuel Clad barrier and a Potential Loss of Containment.

Page 87 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases This IC addresses the inability to restore and maintain reactor vessel level above tile 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 toss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS 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 reqttired.

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 (Table C-2, Contciin,nei,t Cli al/cage Jiiclicc,tion,v).

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 detennine if core uncovery has actually occutTed (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.

Tile inability to monitor RCS level may be caused by instmmentation and/or power failures, or water level dropping below the range of available instnimentation. 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 ttley are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 9 1-283, Evaluation of Shutdowtl and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Conmiercial NucLear Power Plants in the United States; and NUMARC 9 1-06, Guidelines for Industry Actions to Assess Shutdown Management.

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-l77, (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Cotitainment Closctre Page 88 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

5. FSAR. Section 6.2.5

6. FSAR. Table 7A-3 (Sheet 31)

7. Technical Specifications 3.3.3

8. OTN-GS-00001, Containment Hydrogen Control System

9. (ale No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID TiOl & T102

10. NEI 99-0 1, (Si Page 89 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Maiftinclion Subcategory: 1 RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1.2 General Emergency RCS level cannot he monitored for> 30 II1. (Note I)

AND Core uncovery is indicated by any of the following:

UNPLANNED increase in any Table C-I sump/tank level of sufficient magnitude to indicate core uncovery.

Manipulator crane radiation monitor SD-RE-41 > 10,000 mRihr.

Erratic Source Range Monitor indication.

AND Any Containment Challenge indication, Table C-2.

Nuic I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely he exceeded.

Note 6: II [ONTAI MENT CLOSURE is re-established prior to exceeding the 3D-minute time limit, declaration ofa General Emergency is not reqwred.

Table C-I Sumps I Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)

Containment hydrogen concentration 4%

Unplanned rise in Containment pressure MODE Applicability:

5 - Cold Shutdown, 6 Refueling Page 90 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical 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 batTier to fission product release under shtitdown conditions.

As applied to Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.

RCS INTACT- The RCS should be considered intact when the RCS presstire boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

UNPLANNED A parameter change or an event that is not 1) the resttlt 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 ii nkno wn.

Basis:

In Cold Shutdown MODE. the RCS will normally be intact and standard RCS level monitoring means are available.

In the Refueling MODE, the RCS is not intact and RPV Level may be monitored by different means, including the ability to monitor Level visually.

In this EAL. all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-l). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances.

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 RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may he occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref 1, 2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref 3).

The Reactor Vessel inventory loss may be detected by the manipulator crane radiation monitor or erratic Source Range Monitor indication. As water level in the Reactor Vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in up-scaled manipulator crane radiation monitor (SD-RE-41) indication (ref. 4. 5, 6).

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 (ref. 7, 8).

Three conditions are associated with a challenge to Containment integrity:

1. CONTAINMENT CLOSURE not established The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 1 5). If contaimnent closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

Page 91 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

2. Containment hydrogen 4°/b The 4% hydrogen concentration threshold is generally considered the tower limit for hydrogen deflagrations. Callaway is equipped with a Hydrogeti Control System (FICS) which serves to limit or reduce combustible gas concentrations in the Containment. The HCS is an engineered safety feature with redundant hydrogen recombiners, hydrogen mixing system, hydrogen monitoring subsystem, and a backup hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref. 9). Two Containment hydrogen monitors (GS Al-lO and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and ERF IS (ref. 10, II). The hydrogen monitors require a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 12, 13).

3. UNPLANNED rise in Containment pressure An unplanned pressure rise in containment while in Cold Shutdown or RefueLing MODES can threaten Containment Closure capability and thus Contaimitent potentially cannot be relied upon as a barrier to fission product release (ref. 15).

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 IMMEJENT stibstantial core degradation or meLting with potential for loss of containment integrity. Releases can be reasonabty expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended toss of core decay heat removal and inventory makeup, decay heat wilt cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probabte.

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 bum (i.e., at the lower deflagration limit). A hydrogen bum 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 wilt 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. (Table C2, Conrainnie,it Chci/lenge Infliccltions).

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 occutTed (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.

Page 92 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUIvI 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASIS DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The inability to monitor RCS 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 tan.k 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 RCS.

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 9 1-06, Guidelines for Industry Actions to Assess Shutdown Management.

Catlaway Basis Referetice(s):

1. OTO-BB-00003-R014. Excess RCS Leakage

2. OSP-BB-t)0009, RCS inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. FSAR, Section 12.3.3.4

5. FSAR, Table 12.3-2

6. CaIc. No. HPCI -0701, SD-RE-41 Response to Core Uncovery in Refueling MODE

7. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects. pgs. 2-18, 2-19

8. Nuclear Safety Analysis Center (NSAC), 1980, Analysis of Three Mile Island - Unit 2 Accident, NSAC-l

9. FSAR. Section 6.2.5

10. FSAR, Table 7A-3 (Sheet 31)

11. Technical Specifications 3.3.3

12. OTN-GS-00001, Containment Hydrogen Control System

13. CaIc No. 392.2 )CX-95 Catlaway Containment Parameters lOP Action Values, Setpoint ID TIOl & T102

14. OS P-GT-00003, Containment Closure

15. NEI 99-01, CG1 Page 93 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-3, to emergency 4.16KV buses NBO1 and N302 reduced to a single power source for 15 mm. (iVow I)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-3 AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

EDG NEO1

EDG NEO2 MODE Applicability:

5 - Cold Shutdown, 6 Refueling, D Deluded 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 IOCfR5O.2):

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

I. 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.

Page 94 of 239 INFORMATION USE

EIP-ZZ-t)0 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Basis:

for emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses and can be aligned within 15 minutes, whether or not the buses are powered from it.

The criteria for Standby Alignment is that the source can he supplying the station with power within 1 5 minutes. Obviously the Main Transformer could not be aligned for backfeed in 15 mintites during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 1 5 minutes, and credit could be taken for it. The same applies for the Alternate Energy Power System (AEPS). Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby. If AEPS cannot be aligned to a bus within 1 5 minutes, then it is not considered a capable AC power source.

The condition indicated by this EAL is the degradation of the offsite and onsite potver sources such that any additional single failure would result in a loss of alt AC power to the emergency buses.

4.16KV buses NBO1 and NBO2 are the emergency (essential) buses. NBO1 supplies power to Load Group 1 (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (YelLow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBOI and the other source is from the startup transformer XMRO1 via ESF LTC transformer XNBO2. Transformer XNBO1 is the normal supply to bus NBOI; XNBO2 is the normal supply to bus NBO2 (ref. 1, 2, 3).

In addition, NB01 and N302 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. I).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAO 1 and unit auxiliary transformer XMRO2. This is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Stipply (AEPS). AEPS consists of Co-op Power or AEPS Diesel Generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

This cold condition EAL is equivalent to the hot condition EAL SAil.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a toss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may he powering one, or more than one, train of safety-related equipment.

When in the Cold Shutdown, reftieling, or defcieled 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.

Page 95 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 faittire 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 toss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-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.

Callaway Basis Reference(s):

I. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda, Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R0l 1, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, CU2 Page 96 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Reftteling System Malfunction Subcategory: 2 Loss of Emergency AC Potter Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability. Table C-3, to emergency 4.16KV btises NBO I and NBO2 for 15 mm. (Vow 1)

AotL I: The Emergency Coordinator shotild declare the e ent promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-3 AC Power Sources Offs ite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

EDGNEO1

EDG NEO2 MODE Applicability:

5 Cold Shutdown. 6 Refueling, D Defueled Basis:

For emergency classification purposes, capability means that an offsite AC power soul-ce(s) is available to the emergency buses, and is aligned within 1 5 minutes.

If AlPS is not aligned to a btis within 1 5 minutes. then it is not considered a capable AC power source.

The criteria for Standby Alignment is that the source can be supplying the station with power within 15 minutes. Obviously the Main Transformer could not be aligned for backfeed in 1 5 minutes during normal power operations. But, in an outage, and if already aligned for backfeed. the Main Tratisfonner could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AlPS is aligned in standby.

The emergency 4. 1 6KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB0l and NBO2 (ref. 1).

Page 97 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 4.16KV buses NBOI and NBO2 ai-e the etriergency (essential) btises. NBO1 supplies power to Load Group 1 (Red Train) safety related toads and NBO2 stipplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBO1 anct the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBO1 is the normal supply to bus NBO1; XNBO2 is the normal supply to bus NBO2 (ref. 1, 2, 3).

In addition, NBOI and NBO2 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. This is only done during Cold Shutdown unless nuclear safety considerations require it to be done dttrmg hot shutdown when no other power sottrces are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AEPS consists of Co-op Power or AlPS diesel generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SSI.l.

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.

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 emergency bus to service. Additional time is available dtte 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 iower losses.

Escalation of the emergency classification level would be via IC CS I or RS I.

Callaway Basis Reference(s):

I. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda, Section 8.2. 1 .2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R01 1, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, CA2 Page 98 of 239 INFORMATION USE

EIP-ZZ-001t)1 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subeategory: 3 RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200°f.

(Aote 10)

Note 10: Begin monitoring hot condition EALS concurrently for any new event or condition not related to the loss of decay heat removal.

MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

RCS INTACT- The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support naturat circulation cooling.

tii\rpLAV,VED A parameter change or an event that is not 1) the restilt of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification Cold Shutdown ternperawre limit (200°F, ref. 1). These include core exit thermocouples (T/Cs) and Wide Range hot leg temperature indications. PLant computer screens are available for monitoring heatup and cooldown (e.g.. MODE3, KEATU, COOLD, MODE4, ACCUM, and RHR), The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 2. 3, 4).

In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should RCS level indication be subsequently lost.

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification Cold Shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency 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.

Page 99 of 239 INFOR1tIATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technica] Bases This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the Cold Shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the Level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Callaway Basis Reference(s):

1. CallawayTecimical Specifications, Table 1.1-1

2. OTG-ZZ-0000l, Plant Heamp Cold Shutdown to Hot Standby

3. OSP-BB-00007, RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

4. FSAR, Section 7.2.2.3.2

5. NEI 99-0 1, CU3 Page 100 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subeategory: 3 RCS Temperature Initiating Condition: UNPLANNED increase in RCS tempetatttre EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication fork 15 mm.

(Aote I) iVote it 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:

Reactor Vessel water level is normally monitored using the following instruments (ref. 2):

RCS Loop level indications):

Indicators on RLO 1 8:

BB LI-53A, RCS (LOOP 1) HOT LEG LEV RB LI-53B, RCS (LOOP 4) HOT LEG LEV

Computer points:

BBLOO53A, RCS LOOP I HOT LEG LEVEL BBLOO53B, RCS LOOP 4 HOT LEG LEVEL BBLO53BB, RCS LOOP LEVEL - CTMT VENTED

RVUS BB LI-131 1, BR LI-1312, BB LI-l321, and BB L1-1322 (ifin service) (ref 3,4)

Visual observation (if vessel head is removed) (ref. 5)

The Plant Computer System Display called Refuel Level Indications (attn on code RLI) is available to assist in monitoring important parameters cniciat to RCS draining operations (ref. 3).

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification Cold Shutdown temperature limit (200°F, ref 1). These include core exit thermocouples (T/Cs) and Wide Range hot leg temperature indications. Plant computer screens are available for monitoring heatup and cooldown (e.g., MODE3, HEATU, COOLD, MODE4, ACCUM, and RHR). The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 6, 7, 8).

Page 101 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the Level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not estabLished during this event, the Emergency 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 RCS conditions and operators wotild be tinable 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.

Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

Callaway Basis Reference(s):

I. Callaway Technical Specifications Table I. 1-I

2. OOA-BB-00003, Refuel Level Indications

3. OTN-BB-00002, Reactor Coolant System Draining

4. FSAR, Section 18.2.13.2

5. OTS-KE-t)00l8, Draining the Refueling Pool

6. OTG-ZZ-0000l, Plant Heatup Cold Shutdown to Hot Standby

7. OSP-BB-00007, RCS 1-leatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

8. FSAR, Section 7.2.2.3.2

9. NEI 99-01, CU3 Page 102 of 239 INFORMATION USE

EIP-ZZ-00l0I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 3 RCS Temperature Initiating Condition: Inability to maintain plant in Cold Shutdown F AL:

CA3.1 Alert UNPLANNED increase in RCS temperature to > 200°F for> Table C-4 duration. (Notes 1. 10)

OR UNPLANNED RCS pressure increase > 10 psig. (This EAL does nor app/v chiuing irater-soM p/out conditions)

Aotc 1: The Emergency Coordinator shoLild declare the event promptly upon determining that the applicable time has been exceeded, or will likely he exceeded Notc 10: Begin monitoring hot condition EALS concurrently for any new eent or condition not related to the loss of decay heat removal.

Table C-4 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status Heat-up Duration Status RCS INTACT (but not N/A 60 mm.

REDUCED INVENTORY)

RCS Not INTACT established 20 min.*

OR REDUCED INVENTORY not established 0 mm.

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

MODE Applicability:

5 Cold Shutdown, 6 Refueling Definition(s):

COiVTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional hairier to fission product release tinder shutdown conditions.

As applied to Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.

RCS INTACT- The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle darns). The RCS is capable of being placed in an intact condition by Operator Action. i.e., pressurized to support natural circulation cooling.

Page 103 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases UNPLANNED A pal-am eter 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.

REDUCED INVENTORY- Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is tower than 3 feet below the Reactor Vessel flange (< 64.0 in.).

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification Cold Shutdown temperature limit (200°F, ref. I). These include core exit thermocouples (T/Cs) and Wide Range hot leg temperature indications. Plant computer screens are available for monitoring heatup and cooldown (e.g., MODE3, HEATU, COOLD, MODE4, ACCUM, and RHR). The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 2, 3, 4).

RCS pressure instrument 33 PI-403A is capable of measuring pressure to less than 10 psig (ref. 5).

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when the RCS is intact in MODE 5 or based on time to boil data when in MODE 6 or the RCS is not intact in MODE 5.

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification Cold Shutdown temperature limit when the heat removal function is available DOES NOT warrant a classification.

The RCS 1-leat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS 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 RCS Heat-tip Duration Tlwesholcts table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at rectuced 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 ilTadiated fuel.

The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level wotild be via IC CS 1 or RS 1.

Page 104 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases CaIIawa Basis Reference(s):

1. CallawayTecirnical Specifications. Table li-i

2. OTG-ZZ-00001, Plant Heattip Cold Shutdown to Hot Standby

3. OSP-BB-Ot)t)07. RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

4. FSAR, Section 7.2.2.3.2

5. OTG-ZZ-00006, Plant Cooldown Hot Standby To Cold Shutdown

6. NE! 99-01, CA3 Page 105 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TEChNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 4- Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event

< 107 VDC bus voltage indications on Technical Specification required 125 VDC btises for 15 mm.

(Note I)

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:

The purpose of this EAL is to recognize a toss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The fifteen minute interval is intended to exclude transient or momentary power losses, As used ü; 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 (NKO2 orNKO4) would reqtlire the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification, The vital DC buses are the following 125 VDC Class IE buses (ref. I):

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There are four battery banks (NKI1, NKI2, NKI3 and NKI4) that supplement the output of the battery chargers. They sttpply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Page 106 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Due to the load distribution on each of the 125VDC buses, the four batteries for each bus do not have the same rating. All four of the I25VDC buses supply i;werters for 120 VAC NN bus power as well as control power for various safety related systems. NKO1 and NKO4 supply additional DC loads such as diesel field flashing, breaker control power, main control board power and emergency lighting. These loads are not supplied by the other two buses, NKO2 and NK03. For this reason, batteries NKI1 and NKI4 require additional capacity. Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a loss of AC power (station blackout) (ref. 2, 3, 4).

Minimum DC bus voltage is 107.0 VDC (ref 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):

NK El-I (NKO1)

NK 11-2 (NKO2)

NK E1-3 (NKO3)

NK 11-4 (NKO4)

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.l.

This IC addresses a toss ot 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.

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 CAl or CA3, or an IC in Recognition Category R.

Callaway Basis Reference(s):

1. E-21t)lO(Q), DC Single Line Diagram

2. FSAR, Tables 8.3-1, -2, -3

3. FSAR, Section 8.3.2

4. Calculation NK-10, NK System DC Voltage Drop

5. FSAR, Table 8.3A-l III.B

6. ECA-0.0, Loss of All AC Power

7. NET 99-01, CU4 Page 107 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Maiftinction Subeategory: 5 Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-S onsite communication methods.

OR Loss of all Table C-S CRC communication methods.

OR Loss of all Table C-S NRC communication methods.

Table C-5 Communication Methods System Onsite ORO NRC G aitron ics Plant Radios Plant Emergency Dedicated Phones Plant Telephone System ENS (Red Phone) Line Back-Up Radio System Sentry Notification System MODE Applicability:

5 Co[d Shutdown, 6

- - Refueling, D Deflieled Definition(s):

OEFSITE RESPONSE ORGANIZA TIONS (ORO) The State of Missouri (SEMA/MIAC), Callaway County 911/EOC, Gasconade County 911/EOC, Montgomery County 911/EOC and Osage County 91 1/EOC.

Page 108 of 239 INFORMATION USE

EIP-ZZ-O0l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1, 2).

Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page, one Control Room page and four party lines. Communication between parties within the plant can be easily and quickly established by using the general page channel. Communication between parties in the plant and the Control Room can be easily and quickly established using the Control Room page channel. The party line channel is normally tised after the page call is completed. As many as four party lines may communicate simultaneously. The portion of the PA system connecting the fuel transfer area in the Containment, the spent fuel area and new fuel handling area in the fuel building, and the control room can he isolated from the remainder of the PA system from the control room.

This permits extended use of the fuel handling communications system without disruption to the remainder of the system.

2. Plant Radios A six channel 800 MHZ tmiiked radio system for overall plant site area coverage reaches out as far as the intake structure. This two-way radio system provides communications for operating purposes with plant radio-equipped vehicles and plant hand-held portable radios. These systems are for use during normal operation or during a plant emergency. This radio system is available on the Control Room radio consoles, on the security radio consoles, on the EOF radio console, and the TSC radio console. This system is also in the field monitoring team vehicles and is used to communicate during emergencies.

3. Plant Emergency Dedicated Phones Three independent telephone systems are available for communications between the Emergency Response Facilities: the Technical Assessment Bridge Line, the Dose Assessment Bridge Line and the Emergency Management Bridge Line. Each system operates independently from the other systems and allows for conference calls between the members of that bridge line group

4. Plant telephone system The telephone system consists of digital automatic switchboard (DPBX) equipment and telephone stations. The DPBX is provided with redundant processors for reliability. The telephone stations ate located throughout the power block, in the main control room, in the various btiildings around the site, in the security building, and in the service building where the administrative offices are located.

for emergency use, unlisted telephone numbers are provided for direct access to the outside local ptiblic telephone system. Company provided cell phones ARE considered part of the Plant Telephone System. The FLEX response satellite phones are in place for beyond design basis accidents and ARE NOT considered part of the Plant Telephone System.

5. ENS (Red Phone) line The NRC Emergency Notification System (ENS) is an FTS telephone used for official communications with NRC Headquarters. The NRC Headquarters has the capability to patch into the NRC Regional offices. The primary purpose of this phone is to provide a reliable method for the initial notification of the NRC and to maintain continuous communications with the NRC after initial notification. ENS telephones are located in the Control Room, TSC and EOF.

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EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

6. Back-Up Radio System (BURS)

The Back-up Radio System is a communication link between the Catlaway Plant and offsitc emergency response agencies. The primary use of this system is the back-up notification of offsite agencies and the coordination ofoffsite activities during a radio logic al emergency. The system uses 800 MHz radios. There are radio control base: units in the Plant Control Room. TSC and EOF, as well as each county EOC and the State EOC. The backup to this system is the commercial touchtone telephone system Notifications may also be initiated through the Callaway County/City of fulton EOC via the Security radio.

7. Sentry Notification System A computerized notification system linked between the Callaway Plant, the State Emergency Management Agency and the four (4) EPZ risk counties. It allows the Communicator to fill out a notification form on screen and transmit the data simultaneously. Notifications on Sentry can be initiated from the Control Room, the Emergency Operations facility (EOf), or the Technical Support Center (TSC).

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 watTants 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 third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Callaway Basis Reference(s):

I. Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2

2. FSAR, Section 9.5.2

3. NE199-01,CU5 Page 110 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Reftieling 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-6 hazardous event AND Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(A1ut. I I, /2)

Note II: If the hazardous event only results in VISIBLE DAMAGE, with no indication of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Note 12: This EAL is applicable when a Table C-6, Hazardous Events, causes a LOSS OF SAFETY FUNCtION on a SAFETY SYSTEM required for the current operating MODE.

Table C-6 Hazardous Events

EXPLOSION

FIRE

HIGH WINDS or tornado strike

Internal or external FLOODING event

Seismic event (earthquake)

Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Page 111 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Definition(s):

EXPLOSION- A rapid, violent and catastrophic failure ola piece of equipment due to combustion, chemical reaction or over presstlrization. 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.

fII?E 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 temoval, resuLting in a rise of water level within the room or area.

HIGH WINDS Winds in excess of 40 mph (18 m/s) sustained, or 58 mph (26 mIs) gusting.

LOSS Of SAFETY FUNCTION (LOSf,) A LOSF exists when, assuming no conculTent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of the Safety Function Determination Program, a LOSF may exist when a support system is iNOPERABLE, AND:

a) A required system redundant to the system supported by the iNOPERABLE support system is also INOPERABLE; OR b) A required system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR c) A required system redundant to the support system for the supported systems (a) and (b) above is also fNOPERABLE.

SAFETY SYSTEM A system required for sale plant operation, cooling down the plant and/or placing it in the Cold Shtttdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR5O.2):

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

I. 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 DAMA GE 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:

Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-0000l, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2),

Page 112 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

External flooding may he due to high rainfall. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph. (ref. 4).

Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref 5).

An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.

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 nutst 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 fiom the same SAFETY SYSTEM that has indications of degraded performance commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via ICs CS 1 or RS 1.

Callaway Basis Reference(s):

I. OTO-SG-0000l, Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR, Section 3.4 Water Level (flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations

4. FSAR, Section 3.3.1.1 Design Wind Loadings

5. PSAR. Section 9.5.1 fire Protection System

6. NEI 99-0 1, CM Page 113 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emetgency Action Level Technical Bases Category H Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY tEALs 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 tomados, 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 Gases Non-naturally occurring events that can cause damage to plant facilities and include toxic, colTosive, asphyxiant or flammable gas leaks.

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 predetennined 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 fill 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.

Page 114 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards Subcategory: I Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HULl Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.

OR Notification of a credible security threat directed at the site.

OR A validated notification from the NRC providing information of an aircraft threat.

MODE Applicability:

All Definition(s):

SECU]?ITY CONDITION Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site sectirity, 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.

f-IOSTILE ACTION An act toward Callaway 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 incltides 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 Callaway.

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 security shift supervision is defined as the Security Shift Supervisor.

This EAL is based on the Callaway Plant Security Plan and DBT (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 tinder ICs HAL and HSI.

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.

Page 115 of 239 INfOR1IATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Quatification Plan, Safeguards Contingency Plan.

The first threshold references the Shift Security Supervisor because these are the individuals trained to confirm that a security event is occurring or has occulTed. Training on security event confirmation and classification is controlled due to the nature of Safegtiards and 10 CFR 2.39 information.

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Callaway Plant Security Plan and DBT.

The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (ff00) 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 Callaway Plant Security Plan and DBT (ref I).

Emergency plans and implementing procedures are ptiblic 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 Callaway Plant Security Plan and DBT (ref. I).

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

Callaway Basis Reference(s):

I. Callaway Plant Security Plan our! DBT (Safeguards)

2. EIP-ZZ-SKOO1, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-00002, Plant Security Event Aircraft Threat

5. NE199-01,HUI Page 116 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: H Hazards Subcategory: 1 Security Initiating Conditioti: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HAl.! Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.

OR A validated notification from NRC of an aircraft attack threat within 30 mm. of the site.

tIODE Applicability:

All Definition(s):

HOSTILEACTIOiV- An act toward Callaway 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 destnictive force.

Other acts that satisfy the overall intent may be included. Hostile action should not be construed to incltide acts of civil disobedience or felonious acts that are not part of a concerted attack on Callaway.

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).

OIJ7VFR CONTROLLED r1REA 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:

The security shift supervision is defined as the Security Shift Supervisor.

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 Supervisor 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 NET 03-12, Template for the Security Plan, Training and Qualification Plan. Safeguards Contingency Plan.

Page 117 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION UEVEL TECHNICAL BASES DOCUMENT Attachmetit 1 - Emergency Action Level Technical Bases As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite pi-otective 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 1C does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72.

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA.

The second threshold addresses the threat from the impact of an aircraft on the plant. and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely maimer 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 staWs and size of the plane may be provided by NORAD through the NRC.

tn 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 tCs/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 incofl)orate Security-sensitive information. This includes information that may be advantageous to a potential adversaiy, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway PLant Security Plan and DBT (ref. 1).

Callaway Basis Reference(s):

1. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOO1, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-00002, Plant Security Event Aircraft Threat

5. NE199-Ol,HA1 Page 118 of 239 INFORMATION USE

EIP-ZZ-OO 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards Stibcategory: 1 Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occutTing or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor.

NIODE Applicabitity:

All Definition(s):

HOSTILEACTIOiV- An act toward Callaway 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 Callaway.

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 arottnd the process buildings and is depicted in Drawing 8600-X-88 100 Property-Site Layout. Owner Controlled Area and Sttirounding Area.

Basis:

The security shift supervision is defined as the Security Shift Supervisor.

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 Callaway Plant Security Plan and DBT (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, 4 5).

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.

Page 119 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation olonsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (CR0) 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 fttnctions.

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 adeqtiately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 Cf R 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 infornation shotild be contained in non-public documents such as the Callaway Plant Security Plan and DBT (ref. I).

Escalation of the emergency classification level would be via ICs RGI, RG2, SG1, SG2, FGl, and CG1.

Callaway Basis Reference(s):

I. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKt)01, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. 010-S K-0000 1, Plant Security Event Hostile I ntnision

5. OTO-SK-00002, Plant Security Event - Aircraft Threat

6. NE199-0l,HSI Page 120 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 Seismic Event Initiating Condition: Seismic event greater than OBE level EAL:

HU2.1 Unusual Event Seismic event > OBE as indicated by Seismic Activity, Annunciator 98D.

MODE Applicability:

All Definition(s):

None Basis:

Annunciator 98D. OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold (ref. 4).

Seismic Event verification with external sources should not he necessary during or following an OBE.

Earthqtiakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event.

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.

To avoid inappropriate emergency classification restilting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency like the USGS, National Earthquake Information Center (NEIC) can confirm that an earthquake has occulTed in the area of the plant.

Such confirmation should not, however, precltide 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 Callaway. Alternatively, near real-time seismic activity can be accessed via the NEIC website:

http:/./earthquake.usgs.gov/

Additional actions after EAL declaration:

When the seismic recorder indicates that the OBE has been exceeded, as verified by ETP-SG-00001, the reactor must be shut down and remain shut down until inspection of the facility shows that no damage has been incurred which would jeopardize safe operation of the facility or until such damage is repaired.

Callaway was designed such that, for ground motion less than the OBE, those features of the plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. Any ground motion in excess of this results in an uncertainty as to the extent of the damage which must be resolved before continued operation can be considered safe (ref. 1). Ground motion of this magnitude is unmistakabLy a felt earthquake.

Page 121 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases OTO-SG-00001, Seismic Event provides the guidance for determining if the OBE earthquake threshold is exceeded and any reqitired response actions. (ref. 2)

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, stnictures 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.

Depending upon the plant MODE at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Callaway Basis Reference(s):

1. FSAR Section 37(B).1.1 Design Response Spectra

2. OTO-SG-00001, Seismic Event

3. NEI 99-01, HU2

4. FSAR Table 16.33, Seismic Monitoring Instrumentation Page 122 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.I Unusual Event A tornado strike within the PROTECTED AREA.

MODE Applicabitity:

All Definition(s):

PROTECTED AREA An area encompassed by physical barners and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-$8 100 Propeity-Site Layout, Owner Controlled Area and Surrounding Area.

Basis:

Response actions associated with a tornado onsite is provided in OTO-ZZ-00012 Severe Weather (ref. 1).

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EALCA6.1 orSA9.l.

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.l 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.

Callaway Basis Reference(s):

I. OTO-ZZ-000 12, Severe Weather

2. NEI 99-01, HU3 Page 123 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology 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 JOCFR5O.2):

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

I. The integrity of the reactor coolant pressure boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdo\vn condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Refer to EAL CA6. 1 or SA9. 1 for internal or external 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.

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.

Page 124 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Cailaway Basis Reference(s):

1. IPE Section 3.4.2.3 Results of the Vulnerability Screening

2. NEI 99-01, HU3 Page 125 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.3 tJnusual Evcnt 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):

JMPEDE(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 morniarea (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 irocess buildings and is depicted in Drawing 8600-X-88 100 Property-Site Layout. Owner Controlled Area and Stirrounding Area.

Basis:

As used here, the term offsite is meant to be areas external to the Caltaway 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.

The process of flushing chemicals to the sump at the Circ and Service Water btiilding is in a normal plant operations area, but since it is a planned maintenance activity it is excluded provided the process was controlLed. That means looking at extra ventilation and barrier tape to control access. Then as long as the process by which the gasses are being generated is controlled (not expanding beyond the boundary set), is short in duration, and we are able to monitor the atmosphere at the boundary we should be outside the EAL threshold. If we lose control of the process and as a result we had to evacuate part of the Protected Area, then that could meet the IC for HU3.3.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Callaway Basis Reference(s):

1. NEI 99-01, HU3 Page 126 of 239 INFORMATION USE

EIP-ZZ-t)O101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 Natural or Tecimology Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unustial Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.

tNote 7)

Vote 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 stich an event incltide site FLOODfNG caused by a hurricane, heavy tains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended appty 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.

If ALL access roads to the plant are impassable, and local authorities are no longer clearing roads, this EAL applies.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Callaway Basis Reference(s):

1. NET 99-01, HU3 Page 127 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Itiltiating Condition: FIRE potentially degrading the level of safety of the pLant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 mm. of any of the following FIRE detection indications:

(Note I)

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-I 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-i Fire Areas

Area5

Auxiliary Building

Aux Feedwater Pump Rooms

Containment

Control Building/Communications Corridor

Diesel Generator Building

ESW Pumphouse

Fuel Building

RWST

UHS Cooling Tower 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.

Page 128 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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.

Actual fietd reports must be made within the 15 minute time limit or a classification must be made. If a fire is venfied to be occun-ing by field report, the 15 minute time limit is from the original receipt of the fire detection alarm.

Table H-l Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown, TabLe H-l F ire Areas include those structures containing ftinctions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). The Laundry Decon Facility is NOT part of the Aux Building.

This IC addiesses 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 I 5-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 cotild be a drop in fire main pressure, automatic activation of a suppression system.

etc.

The Shift Manager needs to ask some specific questions to ensure they have the information needed to evaluate the situation.

Is there visible flame?

Are there copious quantities of smoke stilt being generated?

A smoked component (subject to overheating) should show blackened areas or signs that the component itself had been very hot (e.g., paint peeling). It can be expected to generate some lower level of smoke. If there is so much smoke present that entry to inspect the component is not possible without an SCBA, that would probably be an indication that a fire existed and determine if EAL HA5.1 is applicable (MODE 4 only). If a breaker truly stiffered a fault local to the breaker, the damage and fire ball would be sttch that consideration of the Hazardous Event EAL SA9.1 would be recommended, if a required Safety System was affected.

In the case of a fire alarm in Containment, OTA-KC-0 1008 states that at the discretion of the Shift Manager/Operating Supervisor, either:

INSPECT detectors for operation AND INSPECT the Reactor Building for the presence of smokefire, OR

INSPECT other contaimnents parameters available in the Control Room. such as other detection zones, containment temperature or equipment failure, for evidence of a fire.

Other items to monitor would be Containment Radiation Monitors such as GTREOO3 1 and GTREOO32 for loss of flow due to filters plugging. The important thing is to make the initial declaration timely with respect to the time of the initial indication. In all cases, document the indications considered for the decision made.

if indications of failing safety related equipment are attributable to the fire, consider Hazardous Event EAL SA9.1 Page 129 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

Depending upon the plant MODE at the time of the event, escalation of the emergency classification level would be via IC CA6 or 5A9.

Callaway Basis Reference(s):

I. FSAR, Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NEI 99-01, HU4 Page 130 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 Fire Initiating Conditioti: FIRE potentially degrading the level of safety of the plant EAL:

11U4.2 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-i area.

AND The existence of a FIRE is not verified within 30 mm. of alarm receipt.

(Note 1)

Ncjte I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or vill likely be exceeded.

Table H-I Fire Areas

Area5

Auxiliary Building

Aux Feedwater Pump Rooms

Containment

Control Building/Communications Corridor

Diesel Generator Building

ESW Pumphouse

Fuel Building

RWST

UHS Cooling Tower MODE Applicability:

At!

Defttiition(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.

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 mtist 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. I.

Page 131 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Table H-i fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown. Table H-I Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). The Laundry Decon Facility is NOT part of the Aux Building.

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.i 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 walTantedi.

The Shift Manager needs to ask some specific questions to ensure they have the information needed to evaluate the situation.

Is there visible flame?

Are there copious quantities of smoke still being generated?

A smoked component (subject to overheating) should show blackened areas or signs that the component itself had been very hot (e.g., paint peeling). It can be expected to generate some tower level of smoke. If there is so much smoke present that entry to inspect the component is not possible without an SCBA, that would probably be an indication that a fire existed and determine if EAL HAS. 1 is applicable (MODE 4 only). If a breaker truly suffered a fault local to the breaker, the damage and fire ball would be such that consideration of the Hazardous Event EAL SA9.l would be recommended, if a required Safety System was affected.

In the case of a fire alarm in Containment, OTA-KC-0l008 states that at the discretion of the Shift Manager/Operating Supervisor, either:

INSPECT detectors for operation AND INSPECT the Reactor Building for the presence of smoke/fire, OR

INSPECT other contaimuents parameters available in the Control Room, such as other detection zones, containment temperature or equipment failure, for evidence of a fire.

Other items to monitor would be Containment Radiation Monitors such as GTREOO3 1 and GTREOO32 for loss of flow due to filters plugging. The important thing is to make the initial declaration timely with respect to the time of the initial indication. In all cases, document the indications considered for the decision made.

If indications of failing safety related equipment are attributable to the fire, consider Hazardous Event EAL SA9.l Page 132 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis-Related Requirements from 10 Cf R 50 10 CFR 50 Appendix A, Criterion 3 states in part:

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.

10 CFR 50.48 Fire Protection states under (2) (iii) The means to limit fire dama%e to structures, systems, or components important to safety so that the capability to shut down the pant safely is ensured.

NFPA 805 Section 1.3.1 states The Nuclear Safety Goal is to provide reasonable assurance that a fire during any plant operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition.

In addition, NFPA 805 Section 4.2.3.3, 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. As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1 -hour time period.

Depending upon the plant MODE at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

Callaway Basis Reference(s):

1. FSAR, Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NE199-Ol,HU4 Page 133 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 mm. of the initial report, alarm or indication.

(Note I)

Note I: The Emergency Coordinator should declare the event proniptt 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. Sotttces of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT reqtiired if large quantities of smoke and heat are observed.

PRO TECTED AI?EA An area encompassed by physical harriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-8$ 100 Property-Site Layout. Owner Controlled Area and Surrounding Area.

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.

li 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.

Callaway Basis Reference(s):

1. NE199-0l,HU4 Page 134 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting PLant Safety Subcategory: 4 Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the 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 balTiers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88 100 Property-Site Layout, Owner Controlled Area and Surrounding Area.

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 offsitc 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.

Callaway Basis Reference(s):

1. NEI 99-01, HU4 Page 135 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardotts 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 that prohibits or IMPEDES access to EITHER of the following: (Note 5)

North Electrical Penetration Room. (Room 1410)

South Electrical Penetration Room. tRoom 1409)

Note 5: If the eqtlipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

MODE Applicability:

4 - Hot Shutdown Definition(s):

JMPEDE(D) Personnel access to a room oi 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 equq)ment, such as SCBAs, that is not routinely employed).

Basis:

The only rooms/areas external to the Control Room that require access to perform field actions consistent with the above criteria for Callaway are the North and South Electrical Penetration Rooms when in MODE 4 to support isolating SI accumuLators and placing RHR in service for RCS cooldown to Cold Shutdown (ref.

1, 2, 3). The equipment required is:

For SI Accumulators:

NGOIBGF3, FDR BKR TO EPHVS$08A SI AC( A OUT ISO, (Room 1410)

NGO2BGF3, FDR BKR TO EPHV88OB SI ACt B OUT ISO. (Room 1409)

NGO1BGF2, FDR BKR TO EPHV8O8C SI ACC C OUT ISO. (Room 1410)

NGO2BHF2, FDR BKR TO EPHV8SOXD SI ACt D oui ISO. (Room 1409)

For A RHR:

NGO2BCF2, FDR BKR TO BBPVS7O2A RCS LOOP 1 HOT LEG TO RHR PMPS ISO. (Room 1409)

NGO1BEF2, FDR BKR TO EJHV87t)IA A RHR PMP SUCT FROM RCS HOT LEG 1 ISO, (Room 1410) for B RHR:

NGO2BBF3, FDR BKR TO BBPV87O2B RCS LOOP 4 HOT LEG TO RHR PMPS ISO, (Room 1409)

NGOIBDF3, FDR BKR TO EJFIVS7OI B B RHR PMP SUCT FROM RCS HOT LEG 4 ISO. (Room 1410)

Page 136 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEt7EL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 entrv 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 roomlarea is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors incltiding 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 roomiarea (e.g.,

requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if any of the following conditions apply:

The plant is NOT in 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.

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 Ks.

Callaway Basis Reference(s):

1. OTG-ZZ-00006 Addendum 06, Securing Safety Injection Accumulators

2. OTN-EJ-0000l Addendum 3, Placing A RHR Train In Service for RCS Cooldown

3. OTN-EJ-0000l Addendum 4, Placing B RHR Train In Service For RCS Cooldown

4. NET 99-01, AA3 Page 137 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation restilting 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 Panel (ASP).

MODE Applicability:

All Definition(s):

None Basis:

The Shift Manager (SM) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhahitability may he caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. OTO-ZZ-00001 Control Room Inaccessibility, provides the instructions for tripping the unit, and maintaining RCS inventory and 1-Jot Shutdown conditions from outside the Control Room (Ref. 1).

For the purpose of this EAL, the 1 5 minute clock starts after determination that Control Room evacuation is necessary, not when OTO-ZZ-00001 Control Room Inaccessibility, is entered.

Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6. I.

This IC addresses an evacuation of the Control Room that restLlts 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 reqttired 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.

Escalation of the emergency classification level would be via IC HS6.

Callaway Basis Reference(s):

1. OTO-ZZ-00001, Control Room Inaccessibility

2. NEI 99-01, HA6 Page 138 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

11S6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel (ASP).

AND Control of any of the following key safety functions is not re-established within 15 mm.:

(\oft 1)

Reactivity (MODE 1, 2, and 3 only).

Core Cooling.

RCS 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:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 Hot Shutdown. 5 Cold Shutdown, 6 - Refueling Definition(s):

None Basis:

For the purpose of this EAL the 15 minute clock, to re-establish controt of key safety fi.tnctions, starts when the last licensed operator leaves the Control Room.

The Shift Manager (SM) detetmines if the Control Room is uninhabitable and requires evacuation. ControL Room iithabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room. or other life threatening conditions. OTO-ZZ-0000l, Control Room Inaccessibility, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. 1, 2).

The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen minute interval.

Page 139 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timeLy maimer. Primary emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shut down the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).

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 isa precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency 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 orCGl.

Callaway Basis Reference(s):

I. OTO-ZZ-0000l, Control Room Inaccessibility

2. OTS-ZZ-000t)l, Cooldown from Outside the Control Room

3. NEt 99-01, E-IS Page 140 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 liE 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):

None Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. I). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response persoimel are notifIed and instructed to report to their emergency response locations. In this maimer, 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.

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 tinder the emergency classification levet description for an Unusual Event.

Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1, Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1, Shift Manager

3. NEI 99-01, KU7 Page 141 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subeategory: 7 Emergency Coordinatoi- Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:

HA7.J 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 Callaway or its personnel that includes the use ofviolent 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 desi.ictive force.

Other acts that satisfy the overall intent may he 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 Callaway.

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 Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initiaLLy acts in the capacity of the Emergency Coordinator and takes actions as outLined in the Emergency Plan implementing procedures (ref. 2). 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 matmer, 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.

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.

Page 142 of 239 INFORMATION USE

EIP-ZZ-Ot)101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan. section 5.1.1 Shift Manager

3. NE! 99-01, HA7 Page 143 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 Emergency Coordinator Judgment Iiiitiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:

1157.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occttrred 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; Alt Definition(s):

IIOSTILEACJJOiV An act toward Callaway or its personnel that includes the tise 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 desfructive 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 ola concerted attack on Callaway.

Nonterrorism-based EALs should be tisecl to address such activities (i.e., this may incltide violent acts between individuals in the owner controlled area)

Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority fur implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). 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.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that vatrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall tinder the emergency classification level description for a Site Area Emergency.

Page 144 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Techtiical Bases Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1 Shift Manager

3. NEI 99-01, I-IS?

Page 145 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazai-ds 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.I 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 fticility. 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 Callaway 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 he construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Callaway.

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).

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

Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). 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, hut 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.

Releases can reasonably he 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.

Page 146 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1 Shift Manager

3. NE199-01,HG7 Page 147 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Cate2orv S System Malfunction EAL Group: Hot Conditions (RCS temperature> 200°f); EALs in this category are applicable only in one or more hot operating MODES.

Numerous system-related equipment failure events that watrant 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:

Loss of Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooting systems which may be necessary to ensure fission product barrier integrity. This category incLudes loss of onsite and offsite sources for 4.16KV AC emergency 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 wan-ant emergency classification. Losses of indicators are in this stiheategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% 5%-

clad failures) is indicative of ftiel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated presstire piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening ftiel clad, RCS and containment integrity.

Page 148 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

6. RIS Failure This subcategory includes events related to failure of the Reactor Trip System (RTS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RTS to complete a reactor trip comprise a specific set of analyzed events refetied 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 RTS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could catise a threat to fuel clad, RES 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 failure failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.

9. Hazardous Event Affectin2 Safety Systems Various natural and technoLogical events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategoiy.

Page 149 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subeategory: 1 Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:

SUI.1 Unusual Eveiit Loss of all offsite AC power capability, Table S-i, to emergency 4.16KV buses NBOI and NBO2 for > 15 mm.

(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-I AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMROJ via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

EDGNEO1

EDG NEO2 MODE Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

For emergency classification purposes, capability means that one of the Table S-i offsite sources remains available and can be aligned within 15 minutes.

The criteria here is that the source can be supplying the station with power within 1 5 minutes. Obviously the Main Transformer could not be aligned for backfeed in 15 minutes during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby. If AEPS cannot be aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

Page 150 o239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NBOI and NBO2 (ref. 1).

NBO1 supplies power to Load Group I (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBOI and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBO1 is the normal supply to bus NBOI; XNBO2 is the normal supply to bus NBO2 (ref. 1, 2 3).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAO1 and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nucLear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AlPS consists of Co-op Power or AEPS diesel generators. Credit can he taken for this source only if it can he aligned within 15 minutes.

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 emergency buses. This condition represents a potential reduction in the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC SAL Callaway Basis Reference(s):

1. E-2100l(Q) Main Single Line Diagram (Electrical Distiibiition Diagram)

FSAR Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-0000 I-ROIl, Main Step-Up Transformer Backfeed IPTE

5. NE199-0l,SU1 Page 151 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

SAil Alert AC power capability, Table S-I, to emergency 4. 16KV buses NBO 1 and N802 redctced to a single power source for 1 5 mm. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

Note I: The Emergency Coordinator shottld declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-i AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

EDGNEO1

EDGNEO2 MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 Hot Shutdown Definition(s):

SAFETY SYSTEAf 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 IOCFR5O.2):

Those structures, systems and components that are relied upon to remain ftmctional 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.

Page 152 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Eniergency Action Level Technica] Bases Basis:

For emergency classification purposes, capability means that one of the Table S-I offsite sources remains available and can be aLigned within 15 minutes.

If AEPS cannot be aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

The criteria here is that the source can be supplying the station with power within 15 minutes. Obviously the Main Transformer could not be aligned for backfeed in 15 minutes during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Tinied control room actions have shown that Catlaway can supply power from AlPS to the station in approximately 9 minutes, 1fAEPS is aligned in standby.

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NBO1 and NBO2 (ref 1).

NB01 supplies power to Load Group I (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBOI and the other sotirce is from the startup transformer XMRO1 via 1SF LTC transformer XNBO2. Transfonner XNBO1 is the normal supply to bus NBO1; XNBO2 is the normal supply to bus N302 (ref 1. 2 3).

Another method to obtain otisite power is by backfeeding the emergency buses through the main transformer XMAO1 and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Stipply (AlPS). AlPS consists of Co-op Power or AlPS diesel generators. Credit can be taken for this source only if it can be aligned within 15 mmutes.

If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared tinder 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 SUI.

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 tinit 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.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

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

Page 153 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R01 1, Main Step-Up Transformer Backfeed [PTE

5. NE199-01,SA1 Page 154 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Conditioii: Loss of all offsite power and at] onsite AC power to emergency buses for 15 minutes or longer EAL:

SSI.1 Site Area Emergency Loss of all offsite and all onsite AC poerto emergency 4.16KV buses NBOI and NBO2 fork 15 m in.

(No/c 1)

Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely he exceeded.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

for this emergency classification this means NBOI and N302 are deenergized for greater than or equal to 15 minutes.

The 4. 16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses N30 I and NBO2 (ref. 1).

NBO1 stipplies power to Load Group I (Red Train) safety related loads and NBO2 stipplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBOI and the other source is from the startup transformer XMRO1 via ESF LTC transformer XNBO2. Transformer XNBOI is the normaL supply to bus NBOL XNBO2 is the normal suppLy to bus NBO2 (ref. 1,2 3).

In addition. NBOI and NBO2 each have an emergency diesel generator (onsite power stipply) which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref 1).

Another method to obtain offsite PO\VCf is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nticlear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

Additional sources of offsite power are available from diesel generators such as the Altemate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and are aligned within 15 minutes. (ref. 5).

Page 155 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 perfotiimnce 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.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RGI, FGI or SG1.

CaIlawa Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR. Section 8.3.1

4. OTS-MA-0000l-R0l I. Main Step-Up Transformer Backfeed - IPTE

5. ECA-0.0, Loss of All AC Power

6. NE199-0l,SSI Page 156 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 1 Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power to emergency 4.16KV buses NBOI and NBO2 AND EITHER:

Restoration of at least one emergency bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely. (vote 1)

CSFST Core Cooling RED Path conditions met.

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4.16KV emergency buses NBOI and NBO2 either for greater then the Callaway Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1,2) or that has resulted in indications ofan actual loss of adequate core cooling.

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. (ref. 3).

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NBO I and NBO2 (ref. 1).

NBO1 supplies power to Load Group I (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transfonuer A or B via 1SF Load Tap Changing (LTC) transformer )U\1B0 1 and the other source is from the startup transformer XMROY via ESF LTC transformer XNBO2. Transformer XNBOI is the normal supply to bus NBOL XNBO2 is the normal supply to bus NBO2 (ref 4, 5 6).

In addition. NBO 1 and NBO2 each have an emergency diesel generator (onsite power supply) which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 4).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref 7).

Page 157 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Additional sources otoffsite power are available from diesel generators such as the Alternate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AlPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and it can be aligned within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . (ref. 8).

Four hours is the station blackout coping time (ref 1, 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 of fission product balTiers and degraded ability to monitor fission product barriers. Indication of continuing core cooLing degradation is manifested by CSFST Core Cooling RED PATI-l conditions being met (ref. 3).

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, contaimuent 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 prodctct 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 fG 1.

This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot he restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation.

Mitigation actions with a low probability olsuccess should not be tised as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare Ihr, and implement, protective actions for the public.

The EAL will also reqttire a General Emergency declatation if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Callaway Basis Reference(s):

1. FSAR, Section 8.3A.5

2. BO-Ol, Station Blackout (SBO) Coping Duration, sh 1

3. CSF-l, Critical Safety Function Status Trees (CSFST) Figure 2, Core Cooling

4. E-2l001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

5. FSAR, Site Addenda Section 8.2.1.2 6 FSAR, Section 8.3.1

7. OTS-MA-0000l-R0l 1, Main Step-Up Transformer Backfeed - IPTE

8. ECA-0.0, Loss of All AC Power

9. NEI 99-01, SG1 Page 158 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency 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 to emergency 4.16KV buses NBOI and NBO2 fork 15 mm.

AND Loss of at! 125 VDC power based on battery bus voltage indications < 107 VDC on all vital DC btises NKOI. NKO3 (Division 1) and NKO2. NKO4 (Division 2) for> 15 miii.

(\otc 1)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will liLelv be exceeded.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby. 4 Hot Shutdown Definition(s):

None Basis:

For this emergency classificatioti this means NBO I and NBO2 are deenergized for greater than or equal to 1 5 minutes.

This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4.16KV emergency buses NBO I and NBO2 for greater than 15 mituites in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiictii.

Ttie 4. 16KV AC System provides the power reqtlirernents for operation atid safe shutdown of the plant. The essential switchgear are buses NBOI and NBO2 (ref. 1).

NBO I supplies power to Load Group I (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNBOI and the other source is from the startup transformer XMRO1 via 1Sf LTC transformer XThJBO2. Transformer XNBOI is the normal supply to bus NBO1; XNBO2 is the nonnal supply to bus NBO2 (ref. 1, 2 3).

In addition, NBO I and NBO2 each have an emergency diesel generator (onsite power supply) which supply eLectrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 1).

Page 159 of 239 INfORIIATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAO1 and unit auxiliary transfomier XMRO2. However, this is onLy done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

Additional sources of offsite power are available from dieseL generators such as the Alternate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AlPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and it can be aligned within 15 minutes. (ref. 5).

The vital DC buses are the following 125 VDC Class 1E buses (ref. 6):

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There are four battery banks, (NK1I, NKI2, NKI3 and NK14) that supplement the output of the battery chargers. They supply DC power to the ctistribution buses when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Due to the load distribution on each of the 125VDC buses, the four battery banks for each bus do not have the same rating. Alt four of the 125 VDC buses supply inverters for 120 VAC NN bus power as well as control power for various safety related systems. NKOI and NKO4 supply additional DC loads such as diesel field flashing, breaker control power. main controL board power and emergency lighting. These loads are not supplied by the other two buses, NKO2 and NKO3. For this reason, batteries NK1 1 and NK14 require additional capacity. Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a loss of AC power (station blackout) (ref. 8, 9, 10).

Minimum DC bus voltage is 107.0 VDC (ref. 9, 10). Bus voltage may be obtained from the following instruments (ref. 6):

NKEI-l(NKO1)

NK EI-2 (NKO2)

NK E1-3 (NKO3)

NK EI-4 (NKO4)

This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A toss of all emergency AC power compromises the performance of aLl SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removaUpressure control, spent fuel heat removal and the ultimate heat sink. A toss of vitaL DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency 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.

Page 160 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases CalIaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R0l 1, Main Step-Up Transformer Backfeed IPTE

5. ECA-0.0, Loss of All AC Power

6. E-21010(Q) DC Single Line Diagram

7. FSAR. Tables 8.3-1, -2, -3 S. FSAR. Section 8.3.2

9. Calculation NK-10, NK System DC Voltage Drop

10. FSAR, Table 8.3A-l III.B

11. NEI 99-01, SG8 Page 161 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: S System MaLfunction Subcategory: 2 Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:

SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications < 107 VDC on all vital DC buses NKO1, NKO3 (Division 1) and NKO2, NKO4 (Division 2) fork 15 mm.

(iVotc 1)

Notc I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely he exceeded.

MODE Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The vital DC buses are the following 125 VDC Class lE buses (ref. 1):

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There are four battery banks, (NKI 1, NK12, NKI3 and NK14) that stipplement the output of the battery chargers. They supply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Dtie to the load distribution on each of the 125VDC buses, the four battery banks for each bus do not have the same rating. All four of the 125 VDC buses supply inverters for 12OVAC NN bus power as well as control power for various safety related systems. NKO1 and NKO4 supply additional DC loads sttch as diesel fieLd flashing, breaker control power, main controL board power and emergency lighting. These loads are not supplied by the other two buses, NKO2 and NKO3. For this reason, batteries NK1 I and NKI4 require additional capacity. Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a Loss of AC power (station bLackout) (ref. 2, 3, 4).

Page 162 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Actioii Level Technical Bases Minimum DC bus voltage is 107.0 VDC (ref. 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):

NKEI-1 (NKO1)

NK E1-2 (NKO2)

NK 11-3 (NKO3)

NK EI-4 (NKO4)

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 tosses.

Escalation of the emergency classification level would be via ICs RG1, FGI or SGI.

Callaway Basis Reference(s):

1. 1-21010(Q) DC Single Line Diagram

2. FSAR. Tables 8.3-1. -2, -3

3. FSAR. Section 8.3.2

4. Calculation NK-l0, NK System DC Voltage Drop

5. FSAR. Table 8.3A-l IIl.B

6. ECA-0.0, Loss of All AC Power

7. NIl 99-01, SS8 Page 163 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 3 Loss of Control Room Indications Initiating Conditioti: UNPLANNED loss of Control Room indications for 15 mintttes 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> I 5 mm.

(Note 1)

Notc 1: The Emergency Coordinator should declare the event promptly upon determlnLng that time limit has been exceeded, or will likely be exceeded.

Table S-2 Safety System Parameters

Reactor power

RCS level

RCS pressure

Core Exit T/C temperature

Level in at least one SIG

Auxiliary or emergency feedwater flow in at least one S/G MODE Applicability:

- Power Operation, 2 - Stat-tup, 3 - Hot Statidby, 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-2 are monitored in the Control Room through a combination of hard controL panel indicators as well as computer based information systems. The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of nonnal Control Room indicators (ref. 1, 2).

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, or any loss of monitoring capabilities of feedwater flow to ALL Steam Generators.

Page 164 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases An event involving a loss of plant indications, atmunciators and/or display systems is evaluated in accordance with 10 Cf R 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, coi-e cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. for example. if the value for reactor vessel level cannot be determined from the indications and recorders on a main control hoard, 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 ctassiflcation level would be via IC SA3.

Callaway Basis Reference(s):

1. FSAR, Section 7.5 Safety-Related Display Instrumentation

2. OTO-RJ-0000l, Loss of Plant Computer

3. NE199-0l,SU2 Page 165 of 239 INFORMATION USE

HP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 3 Loss of Control Room Indications initiating Condition: UNPLANNED toss of Control Room indications for 1 5 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 mm. (Note I)

AND Aiiy significant transient is in progress, Table S-3.

Note i. The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table 5-2 Safety System Parameters

Reactor power

RCS level

RCS pressure

Core Exit T/C temperature

Level in at least one S/G

Auxiliary or emergency feedwater flow in at least one S/G Table S-3 Significant Transients (Automatically or manually initiated)

Reactor trip

Run back 25% thermal power

Electrical load rejection > 25% electrical load

ECCS actuation MODE Applicability:

- 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.

Page 166 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Basis:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computer. which displays SPDS reqtiired 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-3 and include response to automatic or manually initiated functions such as reactor trips, ninbacks involving greater than or equal to 25% thermal power change, electrical load rejections of greater than 25% full electrical load or ECCS (SI) injection actttations.

This IC addresses the diffictilty 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 detenninecl from within the Control Room. This situation would reqtiire 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 procedtires 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 RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for 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 FSI or IC RS1.

Callaway Basis Reference(s):

I. FSAR, Section 7.5 Safety-Related Display Instrumentation

2. QTO-RJ-0000l, Loss of Plant Computer

3. NET 99-01 SA2 Page 167 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 4 RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Sample analysis indicates RCS activity> Technical Specification 3.4.16 limits (listed helow):

>60 p.Ci/gm Dose EquivaLent 1-131.

OR

> 1.0 iCi/gm Dose Equivalent 1-131 for a > 48 hr continuous period.

OR

> 225 pCi/gm Dose Equivalent Xe-133 for a >48 hr continuous period.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 - I-lot Shutdown Definition(s):

None Basis:

This EAL should be entered when the Shutdown Action Statement for Tech Spec 3.4.16 is applied. These values are:

>60 pCi/gm Dose Equivalent 1-13 1.

OR

> 1.0 jiCi/gm Dose Equivalent 1-131 for a > 48 hr continuocts period.

OR

> 225 .tCi/grn Dose Equivalent Xe-133 for a >48 hr continuous period.

(ref 1, 2)

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technicat 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 fAI or the Recognition Category R ICs.

Callaway Basis Referciice(s):

1. Callaway Technical Specifications 3.4. 16 RCS Specific Activity

2. OTO-BB-00005, High Coolant Activity

3. NEI 99-01, SU3 Page 168 of 239 INFORMATION USE

EIP-ZZ-00I0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfitnetion Subcategory: 5 RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage> 10 gpm for? 15 mm.

OR RCS identi fled leakage> 25 gpm for? 15 mm.

OR Leakage from the RCS to a location outside containment > 25 gprn for? 15 mm.

(Note 1)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - I-lot Standby, 4 - Hot Shutdown Definition(s):

N one Basis:

Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage. The Personal Computer (PC) is preIened method of calculating RCS leak rate. When the PC is used, plant status information and all calculations are generated by the Plant Process Computer. When the PC software is not available, procedural guidance is available to perform the manual RCS inventory balance (ref. 1).

Identified leakage includes

Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water leakoff). that is captured and conducted to collection systems or a sump or collecting tank, or

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 presstire boundary leakage, or

RCS leakage through a steam generator to the secondary system (ref. 2).

Unidentified leakage is all leakage (except RCP seal water teakoff) that is not identified leakage (ref. 2).

Pressure Boundary leakage is leakage (except SG tube leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2)

Page 169 of 239 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Ret. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases RCS leakage outside of the containment that is not considered identified or unidentified Leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Component Cooling Water, or systems that directly see RCS pressure outside containment such as Chemical & Volume Control System, Nuclear Sampling system and Residual Heat Removal system (when in the shutdown cooling MODE) (ref. 3, 4)

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA 1 .1.

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the Level of safety of the plant.

The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage, pressure boundary leakage or identified Leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, 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 caletilation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The reLease of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would he required if a mass Loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relict valve sticks open and the line flow cannot be isolated).

The I 5minute threshold duration allows stifficient time for prompt operator actions to isolate the leakage, if possible. If the leak is isolated, the RCS barrier was never lost.

Escalation of the emergency classification level wouLd be via ICs of Recognition Category Ror F.

Callaway Basis Reference(s):

I. OSP-BB-00009, RCS Inventory Balance

2. Callaway Technical Specifications, Definitions Section 1.1

3. FSAR, Section 5.2.5.2.1 hitersystem Leakage

4. OTO-BB-00003-R0l4. Excess RCS Leakage

5. NE! 99-01, SU4 Page 170 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfttnction Subcategory: 6 RTS 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 RTS setpoint is exceeded.

AND A subsequent automatic trip or manuaL trip action taken at the reactor control consoles (SB-HS-1 or SB-HS-42) is sttccess Oil in shutting down the reactor as indicated by reactor power < 5%.

(Note 8)

Vote 8: A manual trip action is any operator action, or set of actions. which causes the control rods to he rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injecuon 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 acti.tation of the automatic Reactor Trip System (RTS) trip function. A reactor trip is automatically initiated by the RTS when certain continuously monitored parameters exceed predetermined setpoints (ref I).

For the purposes of emergency classification, successful manual trip actioiis are those which can be quickly performed from the reactor control console; SB-uS-i on Panel RLOO3 01 SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by use of other trip actions, such as opening PGI9 and PG2O supply breakers, emergency boration or manually driving control rods, do not constitttte a successful manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successftil trip.

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 sottrce 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 RTS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).

Page 171 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Following any automatic RTS trip signal, E-0 (ref. 2) and FR-S.l (ref. 4) prescribe insertion of redundant manual trip signals to back up the automatic RTS trip function and ensure reactor shutdown is achieved.

Even if the first subsequent inantiaL 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).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will automatically initiate auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RTS 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. I.

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 detay 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 RTS 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 thtis 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 shut down 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 shut down 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.

Page 172 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors tnc1udin the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concun-em platit conditiotis. 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 fAl. Absent the plant conditions needed to meet either IC SA6 or FAl, 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., RTS 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 RTS fails to automatically shut down 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 failtire 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.

Note 8 is a generic note applicable to the EALs as approved by the NRC.

Callaway Basis Reference(s):

I. Callaway Technical Specifications, Section 3.3. 1 Reactor Trip System (RTS) Instrumentation

2. 1-0, Reactor Trip or Safety Injection

3. f-0, Critical Safety function Status Trees - Subcriticality

4. FR-SI, Response to Nuclear Power Generation/ATWS

5. FSAR, Section 7.7.1 6 NEI 99-01, SU5 Page 173 of 239 INFORMATION USE

EIP.-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactoi EAL:

SU6.2 Unusual Event A manual trip did not shut dotn the reactor as indicated by reactor potver 5% after any manual trip action was initiated.

AND A subsequent automatic trip or manual trip action taken at the reactor controL consoLes (SB-HS-l or SB-HS-42) is successful in shutting down the reactor as indicated by reactor power < 5%.

(Note $1 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:

Power Operation Definition(s):

None Basis:

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RTS trip setpoint and a subsequent automatic or manual trip is successful in shutting clown the reactor (reactor power < 5%). (ref 1).

For the purposes of emergency classification, successful manual trip actions are those which caii be quickly performed from the reactor cotitrol console; SB-HS-1 on Panel RLOO3 or SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by use of other trip actions, such as opening PG19 and PG2O supply breakers, emergency boration or manually driving control rods, do not constitute a successful manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successful trip.

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 suffucient rod insertion from the trip of RTS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).

Page 174 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Following the failure of any manual trip signal, E-0 (ref. 2) and FR-S.l (ref. 4) prescribe insertion of redundant manual trip signals to back up the RTS trip function and ensure reactor shutdown is achieved.

Even if a subsequent automatic trip signal or the first subsequent manual trip signal inserts all control rods to the fullin position inimediatety after the initial failure of the manual trip, the lowest level of classification that must be dectared is an Unusual Event (ref 4).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will automatically initiate atixiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

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 RTS 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 shut down 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 shut down 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 suiccessM 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 coHdenser, 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 wilt 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 FAI, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Page 175 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Should a reactor trip signal be generated as a result of plant work (e.g., RTS 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 shut down 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.

Note 8 is a generic note applicable to the EALs as approved by the NRC.

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0, Reactor Trip or Safety Injection

3. f-0, Critical Safety Ftinction Status Trees - Subcriticality

4. F R-S.1, Response to Nuclear Power GenerationlATWS

5. FSAR, Section 7.7.1

6. NE199-0l,SU5 Page 176 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 2 RTS 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 (SB-I-IS-I or SB-HS-42) are tiot successful in shutting down the reactor as indicated by reactor power 5%.

(Vote 8)

Aote 8: A manual trip action is any operator action, or set of actions. which catises 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:

- Power Operation Definition(s):

None Basis:

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 has escalated to this EAL.

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB-HS-I on Panel RLOO3 or SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by use of other trip actions, such as opening PGI9 and PG2O supply breakers, emergency boration or manually driving control rods, do not constitute a sttccessfttl manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successful trip.

This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor (reactor power < 5%) followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is prodttcing energy in excess of the heat load for which the safety systems were designed (ref 1).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will atttomatically initiate auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

Page 177 of 239 INFORMATION USE

EIP-ZZ-OOlOl ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 5% rated power is a minimum reading on the power range scale that indicates continued power production.

H 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 3/4 power (ref. 3, 4).

Escalation of this event to a Site Area Emergency would be under EAL SS6. 1 or Emergency Coordinator j udgrnent.

This IC addresses a failtire of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subseqttent operator manual actions taken at the reactor control consoles to shut down the reactor are also unsuccessful. This condition represents an actual or potential stibstantial 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). Tins action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is tinsuccessftil, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the CoHtrol 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 RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS6 or FSI, 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 Categoiy 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.

Note 8 is a generic note applicable to the EALs as approved by the NRC.

Callaway Basis Reference(s):

I. Callaway Technical Specifications, Section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0, Reactor Trip or Safety Injection

3. F-0, Critical Safety function Status Trees - Subcriticality

4. fR-S.l, Response to Nuclear Power GenerationlATWS

5. FSAR, Section 7.7.1

6. NIl 99-0 1, SA5 Page 178 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 2 RTS failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal.

EAL:

SS6.1 Site Area Emergeiicy An automatic or manual trip fulls 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> 50/

AND EITHER:

CSFST Core Cooling RED Path conditions met.

CSFST Heat Sink RED Path conditions met.

MODE Applicability:

- 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.t), 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 RCS barriers.

Reactor shutdown achieved by opening PG19 and PG2O supply breakers, 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 reqtmiring 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 3/4 power (ref. 1, 4).

Page 179 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment I - Emergency Action Level Technical Bases 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 RCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 3).

This IC addresses a failure of the RPS 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 RCS. 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 emel-gency classification resulting from this IC may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs. This is appropriate in that the Recognition Category F ICs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shut down 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 RGI or fG1.

Callaway Basis Reference(s):

1. CSF-l, Critical Safety Function Status Trees Figure 1 Subcriticality

2. CSF-l, Critical Safety function Status Tress Figure 2 Core Cooling

3. CSF-l, Critical Safety Function Status Tress Figure 3 Heat Sink

4. fR-SI, Response to Nuclear Power GenerationlAlWS

5. NE! 99-01, SS5 Page 180 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfttnction Subcategory: 7 Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table 5-4 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 Gaitronics X Plant Radios I X Plant Emergency Dedicated Phones X Plant Telephone System X X X ENS (Red Phone) Line X X Back-Up Radio System Sentry Notification System X NIODE Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definitio ti(s):

OEFSITE RESPONSE ORGANIZATIONS (ORO) The State of Missouri (SEMA/MIAC). Callaway County 911/EOC, Gasconade County 911.(EOC, Montgomery County 911EOC and Osage County 911/EOC.

Page 181 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1, 2).

1. Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page, one Control Room page and four party lines. Communication between parties within the plant can be easily and quickly established by using the general page channel. Comtnunication between parties in the plant and the Control Room can be easily and quickly established using the Control Room page channel. The party tine channel is normally used after the page call is completed. As many as four party lines may communicate simultaneously. The portion of the PA system connecting the fuel transfer area in the Containment, the spent fuel area and new fttel handling area in the fuel building, and the control room can be isolated from the remainder of the PA system from the control room.

This permits extended use of the fuel handling communications system without disruption to the remainder of the system.

2. Plant Radios A six channel $00 MHZ trunked radio system for overall plant site area coverage reaches out as far as the intake structure. This two-way radio system provides communications for operating purposes with plant radio-equipped vehicles and plant hand-held portable radios. These systems are for tise during normal operation or during a plant emergency. This radio system is available on the Control Room radio consoles, on the security radio consoles, on the EOF radio console, and the TSC radio console. This system is also in the field monitoring team vehicles anti is used to communicate during emergencies.

3. Plant Emergency Dedicated Phones Three independent telephone systems are available for communications between the Emergency Response Facilities: the Technical Assessment Bridge Line, the Dose Assessment Bridge Line and the Emergency Management Bridge Line. Each system operates independently from the other systems and allows for conference calls between the members of that bridge line group

4. Plant telephone system The telephone system consists of digital automatic switchboard (DPBX) equipment and telephone stations. The DPBX is provided with redundant processors for reliability. The telephone stations are located throughout the power block, in the main control room, in the various buildings around the site, in the security building, anti in the service building where the administrative offices are located.

For emergency use, unlisted telephone numbers are provided for direct access to the outside local public telephone system. Company provided cell phones ARE considered part of the Plant Telephone System. The FLEX response satellite phones are in place for beyond design basis accidents and ARE NOT considered part of the Plant Telephone System.

5. ENS (Red Phone) line The NRC Emergency Notification System (ENS) is an FTS telephone used for official communications with NRC Headquarters. The NRC Headquarters has the capability to patch into the NRC Regional offices. The primary purpose of this phone is to provide a reliable method for the initial notification of the NRC and to maintain continuous communications with the NRC after initial notification. ENS telephones are located in the Control Room, TSC and EOF.

Page 182 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

6. Back-Up Radio System (BURS)

The Back-up Radio System is a communication link between the Callaway Plant and offsite emergency response agencies. The primary use of this system is the backup notification of effsite agencies and the coordination of offsite activities during a radio logic al emergency. The system uses 800 MHz radios. There ai-e radio control base: units in the Plant Control Room, TSC and EOF, as well as each county EOC and the State EOC. The backup to this system is the commercial touchtone telephone system Notifications may also be initiated through the Callaway County/City of Fulton EOC via the Security radio.

7. Sentry Notification System A computerized notification system linked between the Callaway Plant, the State Emergency Management Agency and the four (4) EPZ risk counties. It allows the Communicator to fill out a notification form on screen and transmit the data simultaneously. Notifications on Sentry can be initiated from the Control Room, the Emergency Operations Facility (EOf), or the Technical Support Center (TSC).

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 ptant or persoirnel safety. this event walTants 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 bemg 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 third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Callaway Basis Reference(s):

1. Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2

2. FSAR, Section 9.5.2

3. NE199-OLSU6 Page 183 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 8 Containment failure Initiating Condition: Failure to isolate containment or loss of contaimuent pressure control.

EAL:

SU8.1 UnusuaL Event Any penetration is not isolated within 15 mm. of a VALID containment isolation signal.

OR Containment pressure> 27 psig with < one full train of containment depressurization equipment operating per design for 15 mm.

(Notes 1, 9)

Note I: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 9: One Containment Spray System train and one Containment Cooling System train comprise one full train of depressurization equipment.

MODE Applicability:

- Power Operation, 2 Startup, 3 - Hot Standby, 4 - I-lot Shutdown Definition(s):

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

Basis:

This EAL addresses 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 of 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.

Page 184 of 239 INFOR1IATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 spray system or containment cooling system) are either lost or performing in a degraded manner.

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the design bases requirement. Each train includes a containment spray pump. spray headers, nozzles, valves, and piping. The refueLing water storage tank (RWST) supplies borated water to the Containment Spray System during the injection phase of operation. In the recirculation MODE of operation, Containment Spray pump suction is transferred from the RWST to the Containment sumps (ref 2).

The Containment Cooling System consists of two trains of Containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement. Each train of two fan units is supplied with cooling water from a separate train of essential service water (ESW). Air is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and otitside the secondary shield in the lower areas of containment. During normal operation, all four fan units may be operating. In post-accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running (ref. 3).

The Containment pressure 1-11-3 setpoint (27 psig, ref. 4, 5, 6) is the pressure at which the equipment should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment Cooling System train (ref 7). Consistent with the design requirement, one friLl train of depressurization equipment is therefore defined to be the availability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

This event wotild escalate to a Site Area Emergency in accordance with IC fSI if there were a concurrent loss or potential loss of either the F tiel Clad or RCS fission product barriers.

Callaway Basis Refereiice(s):

1. FSAR, Section 6.2.2

2. FSAR. Section 6.2.2.1.2.1

3. FSAR. Section 6.2.2.2.2

4. CSf-l, Critical Safety function Status Trees (CSFST) Figure 5, Containment

5. FR-Z. 1. Response to High Containment Pressure

6. Technical Specifications, Table 3.3.2-1

7. Technical Specifications, 33.6.6

8. NET 99-01, SU7 Page 185 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical 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 indication of degraded performance on one train ola SAFETY SYSTEM neected for the current operating MODE AND EITHER:

Event damage has caused indications of degraded perfonnance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(Notes 1/, 12) iVote II: lithe hazardous event only results in VtSIBLE DAMAGE, with no indication of degraded performance to at least one train ofa SAFETY SYSTEM, then this emergency classification is not warranted.

Note 12: This EAL is applicable when a Table S-5 Hazardous Events causes a LOSS OF SAFETY FUNCTION on a SAFETY SYSTEM required for the current operating MODE.

Table S-5 Hazardous Events

EXPLOSION

FIRE

HIGH WINDS or tornado strike

Internal or external FLOODING event

Seismic event (earthquake)

Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 Hot Shutdown Page 186 of 239 INFORMATION USE

E[P-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Definition(s):

EXPLOSION A rapid. violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or over pressurization. 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.

HIGH WIiVDS Winds in excess of 40 mph (18 mis) sustained, or 58 mph (26 m/s) gusting.

LOSS Of SAFETY fUNCTION (LOSF,) A LOSF exists when, assuming no concurrent singLe faiLure, a safety function assumed in the accident analysis cannot be performed. For the purpose of the Safety Function Determination Program, a LOSF may exist when a support system is INOPERABLE. AND:

a) A required system redundant to the system supported by the INOPERABLE support system is also INOPERABLE; OR b) A reqLHred system redundant to the system in turn supported by the INOPERABLE supported system is also INOPERABLE; OR c) A reqtiired system redundant to the support system for the supported systems (a) and (b) above is also INOPERABLE.

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 IOCFR5O.2):

Those structures, systems and components that are relied upon to remain functional dtiring 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:

Annunciator 98D, OBE will illuminate if the seismic instrument detects grout1d motion in excess of the OBE threshold. OTO-SG-0000l, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref 2).

External flooding may be due to high lake level. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

Page 187 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph. (ref. 4).

Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 5).

An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be cLassified under this EAL.

A single FAULTED steam generator would NOT require declaration per this EAL. Technical Specification Bases 3.7.4 explains that two intact Steam Generators are required for cooldown of the RCS and a third Steam Generator is assumed to be RUPTURED. if more than one Steam Generator is FAULTED, then this EAL is applicable.

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 perfonnance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public heatth and safety tiom radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should he 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.

Escalation of the emergency classification level would be via ICs FS1 orRSl.

Callaway Basis Reference(s):

1. OTO-SG-0000 1, Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR, Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations

4. FSAR, Section 3.3.1.1 Design Wind Loadings

5. FSAR, Section 9.5.1 Fire Protection System

6. NE199-0l,5A9 Page 188 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category F Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature> 200°F): 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 pioducts to the environment. The primary fission prodtict barriers are:

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

B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS 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 incltides the containment building and connections tip to and including the outennost containment isolation valves. This barrier also includes the main steam. feedwater, and blowdown line extensions otitside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria cot 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 potentiat loss threshoLds listed in the fission product barrier matrix of Table F-I (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 detennine the appropriate emergency classification level:

Alert:

Any loss or am potential los.s of either fuel Clad or RCS Site Area E,uerency:

Lass or potential loss of any ti io barriers Gen era! Em eren cv:

Loss 0/an) two bcirriers anti loss or potential loss of third jj*

Page 189 of 239 INFORMATION USE

E[P-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The togic used for emergency classification based on fission product barrier monitoring shotild reflect the following considerations:

The fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

Unusual Event ICs associated with RCS and Fuel Clad BalTiers are addressed under System Malfunction ICs.

For accident conditions involving a radiological release, evaluation of the fission product balTier thresholds will need to be performed in conjttnction with dose assessments to ensure cotiect and timely escalation of the emergency classification. F or 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 Callaway design and operaling characteristics.

As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any Location inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a reLief valve is not considered to be RCS leakage.

At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. for example, if the Fuel Clad and RCS Fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product harriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.

Page 190 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: fission Product BaiTier Degradation Subcategory: N/A Initiating Condition: Any loss or any potentiat loss of either Fuel Clad or RCS EAL:

FAI.l Alert Any loss or any potential loss of either fuel Clad or RCS (Table f-i).

MODE Applicability:

- Power Operation, 2 - Startup. 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment 2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. UnLike the Containment balTier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the toss or potential loss of Containment baiTier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL fS 1 .1.

Callaway Basis Reference(s):

I. NE199-0l,FAI Page 191 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: Fission Product BatTier Degradation Subcatcgory: N/A Initiating Condition: Loss or potential loss of any two batTiers EAL:

FSI.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i).

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product baniers. Table F-I (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 otthe following conditions:

One barrier loss and a second barrier loss (i.e., toss - toss)

One baiiier loss and a second barrier potentiaL loss (i.e., loss potential loss)

One barrier potential loss and a second harrier potential loss (i.e., potential loss potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuet Clad and RCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escaLation to a General Emergency is tess imminent.

Callaway Basis Reference(s):

1. NE199-0l,FSl Page 192 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: Fission Product BatTier Degradation Subcategory: N/A Initiating Condition: Loss of an two barriers and loss or potential loss of third harrier EAL:

FG1.1 General Emergency Loss of any two barriers.

AND Loss or potential loss of third harrier (Table F-I).

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment 2) lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each balTier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Contaiimient barriers

Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier

Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier

Loss of Fuel Clad and Containment barriers with potential loss of RCS batTier Caltaway Basis Reference(s):

I. NE199-0l,FGI Page 193 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Ret. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-i lists the thi-eshold 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 colunm is further divided into two columns: one for Loss thresholds and one for Potential Loss thresholds.

The first colunm of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product balTier categories are:

A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / RCS Activity D. CMT Integrity or Bypass I. Emergency Coordinator Judgment Each category occupies a row in Table F-l 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.

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 tfrst Ftiel Clad barrier Loss in Category A would be assigned FC Loss Al, the third Containment barrier Potential Loss in Category C would he assigned CMT P-Loss C.3, etc.

It. a cell in Table FI 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-I 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 barrie;s, the EAL-user first scans down the category column of Table F-l, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the fuel Clad and RCS barriers and a Potential Loss of the Containment batTier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG 1 .1, FS 1 1, and FA 1 .1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Contaimnent 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.

Page 194 of 239 INFORMATION USE

-C z

0 N

N z

(_ .

d z

CI) cI

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Tti res hold:

None.

Page 196 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 197 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 Plant Computer (ref. 1).

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

(altaway Basis Reference(s):

1. CS F-I Critical Satity function Status Trees

2. FR-C. 1 Response to Inadequate Core Coolitig

3. FR-C.2 Response to Degraded Core Cooling

4. NFl 99-01 Inadequate Heat Removal fuel Clad Loss 2.A Page 198 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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- RAGE Path conditions met.

Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core CooLing-ORANGE path incticates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

Callaway Basis Reference(s):

1. CSF- 1 Critical Safety Function Status Trees

2. FR-C. 1 Response to Inadequate Core Cooling

3. FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 199 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 I-Teat sink is required.

Definition(s):

None Basis:

In combination with RCS Potential Loss 8.1, meeting this threshold results in a Site Area Emergency.

Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the 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 Plant Computer (ref. 1).

The phrase and heat sink required precludes the need for classification for conditions in which RCS iresstire 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 I tells the operator to cletenuine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350°f or RCS pressure is greater than 360 psig. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect and place R}IR in service for heat removal. For large LOCA events inside the Containment, the SGs are moot becatise 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 RCS 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.

Callaway Basis Reference(s):

I. CSF-l Critical Safety Function Status Trees Figure 3 Heat Sink

2. fR-H. I Response to Loss of Secondary Heat Sink

3. NET 99-01 Inadequate Heat Removal FueL Clad Loss 2.8 Page 200 of 239 INFORMATION USE

EIP-ZZ-0O 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation> 840 R/hr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s):

None Basis:

Containment radiation monitor readings greater than 840 Rhr (ref. 1) indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 )JCi/cc dose equivalent 1-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within tecimical specifications and are therefore indicative of fuel damage (approximately 5% clad failure depending on core inventory and RCS volume).

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-1 I channel 591) and GT-RE-60 (Panel RM-1 I channel 601). The threshold value of 840 R/hr is the HI-HI (RED) alarm setpoint (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 iCi/gm dose ecjuivalent 1-13 1. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5°/b 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 RCS Barrier Loss threshold C.l since it indicates a loss of both the Fuel Clad Batrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

Callawav Basis Reference(s):

1. EPCI-170l

2. OTA-SP-RMO1I Radiation Monitor Control Panel RM-I 1

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A Page 201 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix aiid Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshoki:

2. Dose equivalent 1-13 1 coolant activity>300 tCi/cc.

Definition(s):

None Basis:

Dose Equivalent Iodine (DEl) is determined by Chemistry procedure CDP-ZZ-08 100, Post Accident Sampling Guidelines (ref. 1).

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 1-1 3 1 concentration is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered tost. (ref. 2).

This threshold indicates that RCS radioactivity concentration is greater than 300 iCi/gm dose equivalent 1-13 1. 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 RCS Activity / Containment Radiation.

Callaway Basis Reference(s):

1. CDP-ZZ-0$ 100 Post Accident Sampling Guidelines

2. NET 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.8 Page 202 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodtict Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

3. CVCS letdown radiation > 2.50E-01 tCi/mI on SJ-RE-01 (016).

Definition(s):

None Basis:

The normal Chemical and Volume Control System (CVCS) charging and letdown flow path allows purification of the reactor coolant and control of the RCS volume while maintaining a continuous feed and bleed flow between the RCS and the CVCS. Reactor coolant is first letdown from the RCS through a regenerative heat exchanger, which minimizes heat losses from the RCS. Additional cooling takes place in a letdown heat exchanger that acts as the heat sink for the system. Downstream of the letdown heat exchanger pressure control valve and ttpstream of the mixed bed demineralizers, the letdown stream passes by radiation monitor SJ-RE-01, which will warn of fission products in the letdown coolant if a fuel element failure occurs. The monitor is located in the Primary Sample Sink Room.

The CVCS letdown monitor SJ-RE-0l provides indication in the Control Room on Panel RM-l 1 channel 016 with a range of 1.7E-03 to l.7E+03 l.tCi/tnL (ref. 2,3). The I-il-HI (RED) alarm is 5E0 + background +

(background x 0.05) (ref. 4) and represents a total fuel clad failure in excess of 1% in 30 minutes (ref. 2, 3).

Five times this alarm setpoint corresponds to approximately 5% fuel clad failure. 5% clad failure is also the basis for the coolant activity and Containment radiation fuel Clad loss thresholds.

Caltawav Basis Reference(s):

1. FSAR Section 9.3.4.2

2. FSAR Table 11.5-1

3. OTA-SP-RMO1 I Radiation Monitor Control Panel RM-l 1

4. HPCI-05-02 Gaseous and Liquid Radiation Monitor Setpoints Rev. 0, Note II

5. NEI 99-0 1 Other Indications Ftiel Clad Loss 5.A Page 203 of 239 INFORMATION USE

EW-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodtict Barrier Loss/Potential Loss [Vlatrix and Bases Barrier: Fuel Clad Category: C. CMI Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 204 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodttct Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. (MT Integrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 205 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

Page 206 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss ltIatrix and Bases Barrier: Fuel Clad Category: F. 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 degtaclation will likely occur within relatively short period of time 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 instntmentation 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.

Callaway Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 207 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: fuel Clad Categoty: I. Emei-gency 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, harrier monitoring capability and dominant accident sequences.

Imminent ban-icr degradation exists if the degradation will likely occur within relatively short period of time based on a projection of dulTent 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 isa loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration 01-0 1tsite monitoring results.

Dominant accident sequences lead to degradation of all fission product harriers and likely entry to the lOPs. The Emergency Coordinator should he mindful of the Loss of AC pveI- (Station Blackotit) 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 ban-icr is potentially lost. The Emergency Coordinator should also consider whether or not to declat-e the barrier potentially lost in the event that barrier status cannot be monitored.

Callaway Basis Reference(s):

1. NEI 99-0 1 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 208 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

UNISOLABLE RCS 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 low pressure < 1 849 psig

Steamline low pressure < 615 psig

Containment high pressure> 3.5 psig

Manual 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). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage. as well as identified leakage. It is also applicable to UNISOLABLE RCS 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.

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 l.A will also be met.

Callaway Basis Reference(s):

1. E-0 Reactor Trip or Safety Injection

2. 1-3 Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss l.A Page 209 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Ttibe Leakage Degradation Threat: Potential Loss TIi reshold:

I. Operation of a standby charging pump is required by EITHER:

UNISOLARLE RCS leakage.

SG tube leakage.

Definition(s):

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

Basis:

This threshold is based on the inability to maintain liquid inventory within the RCS by normal operation of the Chemical and Volume Control System (CVCS). The CVCS includes three charging pumps: one Normal Charging Pump with a design flow capacity of 130 gprn, and two centrifugal charging pumps each with a design flow capacity of 1 50 gpm (ref. 1). Approximately 12 gprn of charging flow bypasses the RCS clue to leakage throtigh the RCP seals; thus, the Normal Charging Pump can deliver 130 gpm 12 gpm = 118 gprn (mundecl to 120 gpm for readability) (ref. 2). A second charging pump being required is indicative of a substantial RCS leak exceeding the capacity of one charging pump (120 gpm) in the normal charging MODE with letdown isolated.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pum be placed in service, indicating a substantial RCS leak exceeding the capacity of one charging pump (120 gpm) in the normal charging MODE with letdown isolated, to restore and maintain pressurizer level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location inside contaimnent, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold l.A will also be met.

Callaway Basis Reference(s):

1. FSAR, Table 9.3-9

2. FSAR, Section 9.3.4 Chemical and Volume Control System

3. E-3 Steam Generator Tube Rupture

4. NEI 99-01, RCS or SG Tube Leakage Reactor Coolant System Potential Loss l.A Page 210 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG lube Leakage Degradation Threat: Potential Loss Threshold:

2. CSFST liitegrity-RED Path conditions met.

Definition(s):

None Basis:

The Potential Loss threshold is defined by the CSFST Reactor Coolant Integrity RED path. CSFST RCS Integrity Red Path plant conditions and associated Pressurized Thermal Shock (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 RCS cooldown under pressure (ref 1, 2).

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock a transient that catises rapid RCS cooldown white the RCS is in MODE 3 or higher (i.e.. hot and pressurized).

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees Figure 4 Integrity and 4a Limit A Curve

2. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition

3. NEI 99-01, RCS or SG Tube Leakage Reactor Coolant System Potential Loss l.B Page 211 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

Page 212 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

Definitioti(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 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 Plant Computer (ref. 1).

The phrase and heat sink required precludes the need for classification for conditions in which RCS 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. lis entered from CSFST Heat Sink-Red. Step I tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350°f or RCS pressure is greater than 360 psig. If these conditions exist, Heat Sink is required, Otherwise, the operator is to either return to the procedure and step in effect and place RHR 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 RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential toss of the RCS 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 RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.

Page 213 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Callaway Basis Reference(s):

1. CSF-1, Critical Safety function Status Trees Figure 3 Heat Sink

2. FR-HI, Response to Loss of Secondary Heat Sink

3. NEt 99-0 1, Inadequate Heat Removal RCS Loss 2.B Page 214 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMI Radiation! RCS Activity Degradation Threat: Loss Iii reshold:

1. Containment radiation> 59 RIhr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s)

None Basis:

Containment radiation monitor readitws greater than 59 Rihr (ref. 1) indicate the release of reactor coolant to the Contaimnent. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-l 1 channel 591) and GT-RE-60 (Panel PJ\f-l I channel 601). The threshold value of 59 RJhr is the 1-lI (YELLOW) alarm setpoint (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. I since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

Callaway Basis Reference(s):

I. EPCI-1701

2. OTA-SP-RMO11, Radiation Monitor Control Panel RM-ll

3. NFl 99-01, CMT Radiation! RCS Activity RCS Loss 3.A Page 215 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation! RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 216 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT liitegrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 217 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

Page 218 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodtict Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: F. Emergency Coordinator Judgnient Degradation Threat: Loss Th resliold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier.

Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS 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 relatively short period of time based on a projection of current safety system performance. The term imrninent 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 incitide instrumentation operability concerns. readings from portable instrumentation and consideration of offsite monitoring results.

Dominant accident sequeices lead to degradation of all fission product harriers and likely entry to the FOPs. The Emergency Coordinator should be mindftd 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 RCS Barrier is lost.

Callawav Basis Reference(s):

1. NFl 99-0 1, Emergency Director Judgment RCS Loss 6.A Page 219 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 RCS barrier.

Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent batTier degradation exists if the degradation will likely occur within relatively short period of time 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 incitide instnmientation operability concerns, readings from portable instmrnentation and consideration of offsite monitoring results.

Dominant accident segttences 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 RCS 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 catmot be monitored.

Callaway Basis Reference(s):

I. NIl 99-01, Emergency Director Judgment RCS Potential Loss 6.A Page 220 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RES or SG Tube Leakage Degradation Threat: Loss Tb reshold:

1. A teaking 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 RES Barrier Potential Loss A.l and Loss Al, respectively. This condition represents a bypass of the containment batTier.

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 priolity condition, the steam generator is stilt 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 reteases oIthis size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the tower bound criteria specified in 1C SU4 for the fuel clad barrier (i.e., RES activity values) and IC SU5 for the RES barrier (i.e.. RCS leak rate vaLues).

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) feedwater 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). If the TDAFP is running and being suppLied by a ruptured steam generator that has not been isolated, this threshold is met. Manual Operator action can NOT be credited.

Page 221 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Steam releases associated with the expected operation of a Steam Generator Atmospheric Steam Dump or Main Steam Safety Valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor ttip 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 eqtial to 25 gpm No classification No classification Unusual Event per Unusual Event per Greater than 25 gpm

- SU5.l Requites operation of a standby charging (makeup) Site Area Emergency Alert per fAI 1 Pu11W per fSI.l (RCS Barrier Potcntial Loss)

Requires an automatic or Site Area Eniereency manual ECCS (SI) actuation Alert per FAll per fSI.I (RCS Bamer Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Callaway Basis Reference(s):

I. E-2, faulted Steam Generator Isolation

2. E-3, Steam Generator Tube Rupture

3. NEI 99-01, RCS or SG Tube Leakage Containment Loss l.A Page 222 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev, 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 223 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodttct Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None.

Page 224 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Prodtict Barrier Loss/Potential Loss Matt-ix 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 mitt (Note I)

Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

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 Plant Computer (ref. I).

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 halTier is also lost.

This threshold addresses any other factors that may be used by the Emergency Coordinator in detenuining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Callawav Basis Reference(s):

1. CSF-1. Critical Safety function Status Trees Figure 2 Core Cooling

2. FR-Cl, Response to Inadequate Core Cooling

3. fR-C.2, Response to Degraded Core Cooling

4. NFl 99-01, Inadequate Heat Removal Containment Potential Loss 2.A Page 225 of 239 INFORIIATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None.

Page 226 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMI RadiationRCS Activity Degradation Threat: Potential Loss Threshold:

1. Containment radiation> 14.0t)0 R/hr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s):

None Basis:

Containment radiation monitor readings greater than 14,000 R/hr (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS batTier and the Fuel Clad barrier.

The readings are higher than that specified for Fuel Clad barrier Loss C. 1 and RCS batTier Loss C. 1.

Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a General Emergency.

Monitors used for this fission prodttct barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-l 1 chatmel 591) and GT-RE-60 (Panel RM-l 1 channel 601) (ref. 2).

The radiation monitor reading colTesponds 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 tised to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe NucLear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major reLease of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a toss of the RCS Barrier and the fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.

Callaway Basis Reference(s):

1. EPCI-1701

2. OTA-SP-RMO1 1, Radiation Monitor Control Panel RM-11

3. NEI 99-01, CMT Radiation / RCS Activity Containment Potential Loss 3.A Page 227 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMI Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required. I AND EITHER:

Containment integrity has been lost based on Emergency Coordinator judgment.

UNISOLABLE pathway from containment to the environment exists.

Definition(s):

UNISOLABLE An open or breached systen; 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 RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a toss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficuLties in cletennining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this thresholct 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 cotild be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS balTier, 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. th this case, the associated steam tine provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS 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.

Page 228 of 239 INFORMATION USE

EIP-ZZ-00 101 ADDENDUI1 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachmetit 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Second Threshold Conditions are such that there is an U1JISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the tetm 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 enviromnent.

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 RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an encLosed system. These releases do not constitute a loss or potential loss of contaimnent but should be evaltiated using the Recognition Categoty R ICs.

The status of the contaimnent barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A. I.

Callaway Basis Reference(s):

1. NE! 99-01, CMI Integrity or Bypass Containment Loss 4.A Page 229 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 RCS Leakage outside of containment.

Definition(s):

None Basis:

ECA-1.2 LOCA Otttside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. PotentiaL RCS leak pathways outside containment include (ref. 1, 2):

Residual Heat Removal

Safety Injection

Chemical & Volume ControL

RCP seals

PZRJRCS Loop sample lines Containment sump, temperature, presstire 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 seqctence). Increases in sump, temperature, pressttre, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.

Unexpected elevated readings and alarms on radiation monitors with detectors otitside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass otitside 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 detemine if RCS mass is being lost outside of the containment.

The sum of the leakage rates of less than or equal to 1 gpm are acceptable outside of containment per Technical Specification. These systems inclttde the recirculation portion of the Containment Spray, Safety Injection, Chemical and Volume Control, and Residtial Heat Removal.

Refer to the middle piping nm of Figure 1. In this simplified example, a teak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage coutd be detected by any of the four monitors depicted in the figure and cause threshold D. 1 to be met as well.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.l to be met.

Page 230 of 239 INFORMATION USE

EIP-ZZ-O0 101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Callawav Basis Reference(s):

1. ECA-I .2. LOCA Outside Containment

2. E-1, Loss of Reactor or Secondary Coolant

3. NEI 99-01. CMI lnterity or Bypass Containment Loss

4. ESP-ZZ-00356, Technical Specification 5.5.2.3 Verification Integrated Leak Rate Requirements for Primary Coolant Sources Outside Containment.

5. Technical Specification 5.5.2.B Page 231 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment IntegriP or Bypass Examples (7 J 4 Thrnchnld-Alrbomc N% (w release ffluent -

.7. Auxiliary Building Ii urn 7 k Mcii tor iiatliay In side Reactor - e t VenQ1

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I ick ;:

-z : 55t r

E :

fL&LJ Cooling Opeivvalve .r.zc Openvalve Pump t

7,:;:F4-RCP Seah Cooling Page 232 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMI Integrity or Bypass Degradation Threat: Potential Loss Threshold:

I 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 48 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref 1).

48 psig is the containment pressure that is expected to occur following a design basis Loss of Coolant Accidetit (LOCA) (ref. 2) and is the pressure used to define CSFST Containment Red Path conditions.

If containment pressure exceeds the pressure that is expected to occur following a design basis Loss of Coolant Accident (LOCA), there exists a potentiat 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 RCS 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 harrier.

Callaway Basis Reference(s):

1. CSF-l, Critical Safety Function Status Trees Containment Figure 5

2. CaIc No. 392.2 XX-95, Callaway Containment Parameters EOP Action Values, Setpoint ID T.03

3. NEI 99-01, CMT Integrity or Bypass Containment Potential Loss 4.A Page 233 of 239 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category; D. CMI Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration 4%.

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 oconstniction and radiolysis of aqueous solution in the core and sump. (ref. 1).

Callaway is equipped with a Hydrogen Control System (l-ICS) which serves to limit or reduce combustible gas concentrations in the Containment. The HCS is an engineered safety feature with redundant hydrogen i-ecombiners, hydrogen mixing system, hydrogen monitoring subsystem, and a backup hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref. 1).

HCS operation is prescribed by EOPs if Containment hydrogen concentration should reach 0.5% by volume (ref. 4). If the Potential Loss threshold is reached or exceeded, the primary means of controlling Containment hydrogen concentration must have failed to perform its design function or has otherwise been inadequate in mitigating the hydrogen generation rate. for either case, continued hydrogen production may yield a flammable hydrogen concentration and a consequent threat to Containment integrity.

To generate such levels of combustible gas, loss of the fuel Clad and RCS barriers must have occun-ed.

With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.

Two Containment hydrogen monitors (GS Al-b and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and ERF IS (ref. 3). The hydrogen monitors require a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 4, 5). If an actttal hydrogen concentration measurement is unavailable, CA-3 (ref. 6) may be used to estimate the Containment atmosphere hydrogen concentration.

The existence of an explosive mixture means, at a minimtim, that the containment atmospheric hydrogen concentration is sufficient to stipport a hydrogen bum (i.e., at the lower deflagration limit). A hydrogen bum 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.

Page 234 of 239 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Callaway Basis Reference(s):

1. FSAR. Section 6.2 Containment Systems

2. FR-Z.4. Response to High Containment Hydi-ogen Concentration

3. FSAR, Table 7A-3 (Sheet 32 Data Sheet 6.4)

4. OTN-GS-00001, Containment Hydrogen Control System

5. CaIc No. 392.2 )O(-95, Callaway Containment Parameters EOP Action Vatues, Setpoint ID TIOl &

Ti 02

6. CA-3, Hydrogen Flammability in Containment

7. NEI 99-01, CMT IntegrIty or Bypass Containment Potential Loss 4.B Page 235 of 239 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 > 27 psig with < one hill train of containment depressurization equipment operating per design for 15 mm.

(Notes 1. 9) iVote 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 9: One Containment Spray System train and one Containment Cooling System train comprise one fctll train of clepressurization equipment.

Definition(s):

None Basis:

The Containment Spray System consists of two separtite trains of equal capacity, each capable of meeting the design bases requirement. Each train includes a containment spray pump, spray headers, nozzles, valves, and piping. The refueling water storage tank (RWST) supplies horated water to the Containment Spray System during the injection phase of operation. In the recirculation MODE of operation, Containment Spray pump suction is transferred from the RWST to the Containment sumps (ref. 2).

The Containment Cooling System consists of two trains of Containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement. Each train of two fan ttnits is supplied with cooling water from a separate train of essential service water (ESW). Air is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield in the lower areas of containment. During normal operation, all four fan units may be operating. In post-accident operation following an actuation signal, the Containment Cooling System fins are designed to start automatically ü slow speed if not already running (ref. 3).

The Containment pressure setpoint (27 psig, ref. 4, 5, 6) is the pressure at which the equipment should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment CooLing System train (ref. 7). Consistent with the design requirement, one full train of depressurization equipment is therefore defined to be the avaiLability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

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 hill 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.

Page 236 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier- Loss/Potential Loss Matrix and Bases Callaway Basis Reference(s):

1. FSAR, Section 6.2.2

2. FSAR, Section 6.2.2.1.2.1

3. FSAR, Section 6.2.2.2.2

4. CSF-1, Critical Safety function Status Trees (CSFST) Figure 5. Containment

5. FR-Z.1, Response to High Containment Pressure

6. Technical Specifications. Table 3,3.2-1

7. Technical SpeciFications, B3.6.6

8. NFl 99-0 1, CMT Integrity or Bypass Containment Potential Loss 4.C Page 237 of 239 INFORMATION USE

EIP-ZZ-00l0l ADDENDUIvI 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. 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 batrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occttr within relatively short period f time 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 monitorinz capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings tlom portable instrumentation and consideration of o ffsite 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 he used by the Emergency Coordinator in determining whether the Containment Batrier is lost.

Callaway Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment PC Loss 6.A Page 238 of 239 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 016 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment 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 Containment barrier.

Definition(s):

None Basis:

The Emergency Coordinatorjudgrnent threshold addresses any other factors relevant to determining if the Primary Containment bamer is potentially lost. Such a determinatton should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent batTier degradation exists if the degradation will likely occur within relatively short period of time 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.

Callawav Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment PC Potential Loss 6.A Page 239 of 239 INfORMATION USE

Attachinent 4 to ULNRC-06433 EIP-ZZ-0O 101 Addendum 1, Emergency Action Level Classification Matrix, Proposed Revision 009 (Clean Copy Information Only)

(8 pages)

Ame ren MISSOURI Callaway Energy Center EIP-ZZ-OO1O1 ADDENDUM 1 EMERGENCY ACTION LEVEL CLASSIFICATION MATRIX MAJOR Revision 009 Page 1 of 8 INFORMATION USE

EIP-ZZ-00101 ADDENDUM I Rev. 009 EMERGENCY ACTION LEVEL CLASSIFiCATION MATRIX TABLE OF CONTENTS Section Page Number 1.0 PURPOSE 3 2.0 SCOPE 3 3.0 PROCEDURE INSTRUCTIONS 3

4.0 REFERENCES

7 4.1. Implementing 7 4.2. Developmental 7 5.0 RECORDS 7 6.0

SUMMARY

OF CHANCES 8 Page 2 of 8 INFORMATION USE

EIP-ZZ-t)0l0l ADDENDUM 1 Rev. 009 EMERGENCY ACTION LEVEL CLASSIFICATION MATRIX 1.0 PURPOSE Provide a method of consistently and accurately classiiing Emergency Action Levels (EAL).

2.0 SCOPE Covers EALs at the Callaway Nticlear Power Plant.

3.0 PROCEDURE INSTRUCTIONS This document is a copy of the actual EAL Classification Matrix wall charts posted in required areas.

The wall charts are designed for f-lot Conditions (RCS > 200°f) and Cold Conditions (RCS 200°F) with the left side of both charts identical for ease of use, as the left side covers All Conditions. The following pages of this document will break the wall chart in to 3 sections for ease of reading within the constraints of a procedure type document. During an Emergency, the actual wall charts should be used.

Page 3 of 8 INFORMATION USE

EIP-ZZ-00I0l ADDFNDUL\I I Rev. 009 SiTE AREA EMERGENCY I ALERT - UNUSUAL EVENT rZEL 15 c°° CTli+/-zZCl 1 5s .1 ij l1 - ,r sq spruostt..ns u:.ottn rr0r0s1,,

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[ Table F-I Fission Product Gamer Matrix j Fuel ClaG IFCI Bsnrer Recolor Coolant Sysreor IRCS) Battier Conrainmenl COlT) Garner Category Loon Potential loss Loss Polenrial toss toss Potential Loss A Z.

I RC3nuiT.ioe - sososerneoeosectt, B

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HOTCONDITIONS(PUS>200 F) irn nt Page 4 of S INFORI\TION (SE

EIP-LZ-flOIOI *\DDENDL\I I Rex. 009 SITE AREA EMERGENCY ALERT UNUSUAL EVENT i.-4 I 2EELrZJLZ I a a .LJLLrrr rTzzLrre _LaL I

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COLD CONDITIONS (RCS 200°F) scurn yc Page 6 of 8 (SE

EIP-ZZ-00101 ADDENDUM 1 Rev. 009

4.0 REFERENCES

4.1. Imp lementin 4.1.1. EIP-ZZ-00 101, Classification of Emergencies 4.1.2. EIP-ZZ-00260. Event Closeout / Plant Recovery 4.1.3. RERP Chapter 4

4. 1 .4. EAL Compat-ison Matrix 4.1.5. EAL Classification Matrix 4.2. Developmental 4.2. 1. N El 99-01. Methodology for Development of Emergency Action Levels 4.12. NUMARC/NESP-007 Rev. 2, Methodology for Development of Emergency Action Levels, Questions and Answers 4.2.3. NRC Regulatory lsstie Summary (RIS) 2003-18. Supplement 2. Use of Nticlear Energy Institute (NEI) 99-0 1, Methodology for Development of Emergency Action Levels Revision 4, Dated January 2003 (December 12, 2005) 4.2.4. NRC Bulletin 2005-02 Emergency Preparedness and Response Actions for Security-Based Events 4.2.5. Callaway Plant Radiological Emergency Response Plan (RERP) 4.2.6. Technical Specifications B.3.9.4 4.2.7. OS P-GT-00003 Containment Closure 4.2.8. APA-ZZ-00 102 EOP/OIO Writers Manual 4.2.9. Procedtire Writing Manual, Cal laway Nuc tear Power Plant 4.2.10. ULNRC-05281, Docket Number 50-483, Callaway Plant Union Electric Co.

Application for Amendment to Facility Operating License NPf-030 Adoption of lndtistry Traveler TSTF-490. May 9, 2006. Attachments 1, 2, 3 and 4 4.2.11. CR 201702763, NOS Insight EP Risk Significant Planning Standard Performance Upper Tier Cause Evaluation Needed 5.0 RECORDS None Page 7 of$ INFORMATION USE

EIP-ZZ-0010l ADDENDUM I Rev. 009 6.0 StIMMARY OF CHANGES Section or Page(s) Description Step Number 5 Old EAL HG I I Deleted HG 1.1 due to the ambiguity of the EAL and that other EALs would cover escalation of a security event.

Reworded the EAL to better address the intent. The new wording reads:

The occurrence of any Table 0-6 (S-5) hazardous event AND Event damage has caused indication of degraded performance on train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a I

EAL SA9 I second train of a SAFETY SYSTEM needed for the current operating 4 EALCA6.l MODE.

6 & Note block on

Event damage has resulted in VISIBLE DAMAGE to a second train of each page a SAFETY SYSTEM needed for the current operating MODE.

I (Notes 77. 72)

And along with the revised EAL text, the following notes were also added:

Note 11: If the hazardous event only results in VISIBLE DAMAGE, with no indication of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Note 12: This EAL is applicable when a Table 0-6 (S-5), Hazardous Events, causes a LOSS OF SAFETY FUNCTION on a SAFETY SYSTEM required for the current operating MODE. -

Page $ of 8 INFORMATION USE

ULNRC-06433, License Amendment Request for Emergency Action Level (EAL) Changes

Contents

Text

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

8.0 REFERENCES

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

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ULNRC-06433, License Amendment Request for Emergency Action Level (EAL) Changes

Text

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

8.0 REFERENCES

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

SUMMARY

4.0 REFERENCES

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