ULNRC-06227, Response to Request for Additional Information Related to License Amendment Request for Emergency Action Level Upgrade Adopting NRC-Endorsed NEI 99-01, Rev 6

Response to Request for Additional Information Related to License Amendment Request for Emergency Action Level Upgrade Adopting NRC-Endorsed NEI 99-01, Rev 6

ML15187A379
Person / Time
Site:	Callaway
Issue date:	07/06/2015
From:	Maglio S Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
TAC MF4945, ULNRC-06227
Download: ML15187A379 (42)

Text

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WAmeren Callaway Plant MISSOURI July 6, 2015 ULNRC-06227 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.47 10 CFR 50.54(q) 10 CFR 50 Appendix E, IV .B.2 10 CFR 50.90 Ladies and Gentlemen:

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

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

RELATED TO LICENSE AMENDMENT REQUEST FOR EMERGENCY ACTION LEVEL (EAL) UPGRADE ADOPTING NRC-ENDORSED NEI 99-01, REVISION 6 (TAC NO. MF4945)

By letter dated October 2, 2014 (ADAMS Accession Number ML14275A435) Ameren Missouri submitted a license amendment request to upgrade the Emergency Action Level scheme associated with the Radiological Emergency Response Plan (RERP) for Callaway Unit 1 by adopting NRC-endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors." During its review, the NRC staff determined that requests for additional information (RAis) were needed to complete its review. Pursuant to electronic correspondence dated May 14, 2015 and a clarification teleconference on May 21, 2015, the NRC transmitted a list ofRAis to Ameren Missouri in electronic form on May 22, 2015, and requested that responses be provided within 45 days.

The responses to the RAis are provided in Attachment 1 to this letter, and supporting documentation is provided in Attachments 3, 4 and 5. Additional proposed changes to the EAL Bases document are listed in Attachment 2. For information, a revised copy of the proposed EAL wall charts is provided in Attachment 6.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • PO Box 620 Fulton, MO 65251 AmerenMissouri.com **************

STARS

• Alliance

ULNRC-06227 July 6, 2015 Page 2 The Callaway Onsite Review Committee has approved the proposed changes to the RERP and its Bases. In addition, in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," Section (b)(l), a copy of this letter is being provided to the designated Missouri State official.

This submittal does not contain new commitments. For any questions concerning this letter, contact Gene Juricic at 573-676-4489 or Pat McKenna at 573-676-8504.

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

Sincerely, Executed on: July 6, 2015 A)aotf A . ~

Scott A. Maglio Manager, Regulatory Affairs JPK/nls Attachments:

1) Response to Request for Additional Information (RAI) Emergency Action Level (EAL)

Scheme Change

2) Summary ofEAL Changes NOT Associated with RAJ Responses

3) Basis for Spent Fuel Storage Cask Dose Rate Limits (Excerpt from Certificate of Compliance No 1040 Amendment No. 0, Appendix A)

4) Basis for Difference in NEI 99-01 Revision 6 and Revision 5 Emergency Action Level Table R-1, "Effluent Monitor Classification Threshold Values"

5) Calculation EPCI-08-01 Rev. 1, "Dose Projection Calculations to Support NEI 99-01 Rev.5 Emergency Action Levels"

6) Callaway NEI 99-01 Revision 6 EAL Wall Charts (Information Only)

ULNRC-06227 July 6, 2015 Page3 cc: Mr. Marc L. Dapas Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Director Division of Spent Fuel Management U.S. Nuclear Regulatory Commission Washington, DC 20555-2738 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. L. John Klos Project Manager, Callaway Plant Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8B 1 Washington, DC 20555-2738 Senior Emergency Preparedness Analyst U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511

ULNRC-06227 July 6, 2015 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4150 International Plaza Suite 820 Fort Worth, TX 76109 (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 D. W. Neterer L. H. Graessle T. E. Herrmann B. L. Cox P. J. McKenna S. A. Maglio T. B. Elwood Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Mr. Robert D. Stout (DNR)

Ms. Leanne Tippett-Mosby (DNR)

Missouri Public Service Commission Mr. Greg Voss, REP Manager (SEMA)

Attachment 1 to ULNRC-06227 Response to Request for Additional Information (RAI)

Emergency Action Level (EAL) Scheme Change 19 Pages

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response Section 4.3, Instrumentation Used for EALs, to CP has confirmed that all setpoints and indications used in the CP EAL NEI 99-01, Revision 6, states Scheme developers scheme are within the calibrated range(s) of the stated instrumentation and that should ensure that specific values used as EAL the resolution of the instrumentation is appropriate for the setpoint/indication.

setpoints are within the calibrated range of the referenced instrumentation. Please confirm that 01 4.3 all setpoints and indications used in the CP EAL scheme are within the calibrated range(s) of the stated instrumentation and that the resolution of the instrumentation is appropriate for the setpoint/indication.

In regards to Section 1, Purpose, of the proposed EAL Technical Basis:

a. Deleted: It should be used to facilitate review of the Callaway EALs and

a. Section 4.6, Basis Document, to NEI 99-01, provide historical documentation for future reference.

Revision 6, states A basis document is an integral part of an emergency classification scheme. The material in this document supports proper emergency classification decision-making by providing informing background and development information in a readily accessible format. It can be referred to in training situations and when making an actual emergency classification, if necessary.

02 1.0 Please revise Section 1 of the proposed EAL Technical Basis to reflect the intent of the EAL Basis Document, as provided in NEI 99-01, Revision 6, and remove the proposed purpose discussion, to facilitate a review of the Callaway EALs, or provide justification for failure to align with NRC endorsed guidance.

Page 1 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response

b. Section 4.6, Basis Document, to NEI 99-01, b. The following has been added to Section 1.0 Introduction of the Technical Revision 6, states Because the information in Bases document:

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

related to maintaining the technical basis document in accordance with 10 CFR 50.54(q) or provide justification for failure to align with NRC endorsed guidance.

c. Section 4.7, EAL/Threshold References to c. The following has been added to Section 1.0 Introduction of the Technical AOP [Abnormal Operating Procedure] and Bases document:

EOP [Emergency Operating Procedure] Additionally, changes to plant AOPs and EOPs that may impact EAL bases Setpoints/Criteria, to NEI 99-01, Revision 6, shall be evaluated in accordance with the provisions of 10 CFR 50.54(q).

states As reflected in the generic guidance, the criteria/values used in several EALs and fission product barrier thresholds may be drawn from a plants AOPs and EOPs. The NRC staff expects that changes to AOPs and EOPs will be evaluated in accordance with the provisions of 10 CFR 50.54(q). Please incorporate information related to screening changes to AOPs or EOPs to determine if an evaluation pursuant to 10 CFR 50.54(q) is required or provide justification for failure to align with NRC endorsed guidance.

Page 2 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response Sections 2.1, Background, and 4.0, Revised the referenced ADAMS Accession No. to ML12326A805.

References, of the proposed EAL Technical Basis reference an incorrect ADAMS Accession 03 2.1 Number (ML110240324). Please verify that the proposed EAL Technical Basis is consistent with NRC endorsed guidance and appropriate ADAMS Accession number is referenced.

For Sections 2.1, Background, and 4.0, Revised the referenced ADAMS Accession No. to ML12326A805.

References, of the proposed EAL Technical 04 2.1 Basis, please provide ADAMS Accession Number that references the endorsed version of NEI 99-01, Revision 6 (ML12326A805).

Section 2.5, Technical Basis Information, of the Separate site-specific and generic bases were identified within the EAL bases proposed EAL Technical Basis includes a Plant- to facilitate NRC review. These bases sections have now been combined into a Specific basis section, in addition to a Generic single bases section for each EAL.

basis section. Considering that the EAL Technical 05 2.5 (Throughout document)

Basis is provided to support proper emergency classification decision making, please explain why a Generic basis section is provided or revise accordingly.

Page 3 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response Section 2.6, Operating Mode Applicability, of the Revised the cited section to read consistent with Section 5.4 of the generic proposed EAL Technical Basis contains a brief guidance:

discussion concerning EAL classification during The mode in effect at the time that an event or condition occurred, and prior mode changes. However, this discussion is not as to any plant or operator response, is the mode that determines whether or clear as that provided in NRC endorsed guidance.

not an IC is applicable. If an event or condition occurs, and results in a mode Please justify the omission of significant portions change before the emergency is declared, the emergency classification level of Section 5.4, Consideration of Mode Changes is still based on the mode that existed at the time that the event or condition During Classification, of NEI 99-01, Revision 6, or 06 2.6 was initiated (and not when it was declared). Once a different mode is revise accordingly.

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.

Page 4 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response In regards to Section 3 of the proposed EAL Technical Basis:

Section 3.1.1, Classification Timeliness, includes Added the cited wording to Section 3.1.1:

a reference to NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power When assessing an EAL that specifies a time duration for the off-normal Plants, but does not include a discussion, as condition, the clock for the EAL time duration runs concurrently with the provided by NEI 99-01, Revision 6, Section 5.2, emergency classification process clock.

Classification Methodology, addressing [w]hen assessing an EAL that specifies a time duration 3.1.1 for the off-normal condition, the 'clock' for the EAL 07 time duration runs concurrently with the 3.1.2 emergency classification process 'clock'. Please justify excluding this information or revise accordingly.

Section 3.1.2, Valid Indications, does not include Added the cited statement to Section 3.1.2:

statement, [t]he validation of indications should The validation of indications should be completed in a manner that supports be completed in a manner that supports timely timely emergency declaration, emergency declaration, as provided by NEI 99-01, Revision 6, Section 5.1, General Considerations. Please justify excluding this information or revise accordingly.

Page 5 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response Appendix B, Definitions, to NEI 99-01, Revision Added the following definitions to Section 5.1:

6, provides definitions for key terms necessary for overall understanding of the NEI 99-01 emergency

• Confinement Boundary, classification scheme. For Section 5.1,

• Emergency Action Level Definitions, please revise accordingly to add

• Emergency Classification Level definitions for the following or justify excluding:

• Fission Product Barrier Threshold

• Initiating Condition.

• CONFINEMENT BOUNDARY, 08 5.0

• EMERGENCY ACTION LEVEL,

• EMERGENCY CLASSIFICATION LEVEL,

• FISSION PRODUCT BARRIER THRESHOLD, and

• INITIATING CONDITION.

In addition, please consider removing one of the two provided definitions for INDEPENDENT Deleted the duplicate ISFSI definition from Section 5.1.

SPENT FUEL STORAGE INSTALLATION to eliminate redundancy.

Section 6.0, Callaway to NEI 99-01 Rev. 6 EAL Cross-Reference, contains the several apparent inconsistencies, as listed below. Please review the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference for accuracy and make corrections as needed.

a. Callaway emergency action level RA2.3 is not a. RA2.3 added to EAL Cross-Reference.

included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference matrix.

09 6.0

b. Callaway emergency action level CG1.1 b. Revised EAL Cross-Reference to cite correct reference.

corresponds to NEI 99-01 Rev. 6 EAL CG1 example 1. The Callaway to NEI 99-01, Revision 6, EAL Cross-Reference indicates example 2.

c. Callaway emergency action level CG1.2 is not c. CG1.2 added to EAL Cross-Reference.

included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference.

Page 6 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483

d. Callaway emergency action level SU 4.1 d. Corrected cited reference.

corresponds to NEI 99-01, Revision 6, EAL SU3 example 2. Callaway to NEI 99-01, Revision 6, EAL Cross-Reference indicates example 1.

e. Deleted SU4.2 from EAL Cross-Reference. Generic SU 3 example 1 is not

e. The Callaway to NEI 99-01, Revision 6, EAL Cross-Reference includes Callaway EAL implemented at Callaway.

SU4.2. EAL SU4.2 could not be located in the Callaway EAL basis document.

f. Callaway emergency action level SU8.1 is not f. Added SU8.1 to EAL Cross-Reference included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference matrix.

g. The Callaway to NEI 99-01, Revision 6, EAL g. Corrected to read SA9.1 Cross-Reference includes Callaway EAL SA8.1. EAL SA8.1 could not be located in the Callaway EAL basis document.

h. Callaway emergency action level SA9.1 is not h. See g above.

included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference matrix.

i. Callaway EAL EU 1.1 shows as IU1.1 on the i. Corrected designation to read E Callaway to NEI 99-01, Revision 6, EAL Cross-Reference. (Note: The E designation is correct.)

Page 7 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL RU1.1, it is not clear how a determination GT-RE-21B, GH-RE-10B & HB-RE18 alarms, Hi-Hi alarm setpoints, and can be made that a 2 X Hi - Hi alarm condition monitor indications are displayed on the RM-11 in the Control Room. It is a exists. Please provide justification that a value of simple matter of multiplying by the Hi-Hi alarm setpoint by 2 to get the EAL 10 RU1.1 two times the alarms identified in Table R-1, indicator value. If the RM-11 Hi-Hi alarm were to come in, the Control room Effluent Monitor Classification Thresholds, can Operator would monitor the parameter for EAL applicability.

be accurately determined in a timely and accurate manner.

For EALs RA1.2, RS1.2, and RG1.2, please Deleted Note 3 applicability to RA1.2, RS1.2 and RG1.2.

explain why the proposed Note 3, which relates to RA1.2, RS1.2, 11 effluent flow past an effluent monitor, should be RG1.2 included for EALs that are based on dose assessments or revise accordingly.

The NEI 99-01 Revision 5 Table R-1 calculation is documented in EPCI-08-01 For EALS RA1.1, RS1.1, and RG1.1, there was a (refer to Attachment 5). The software program MAGNEM was used. Selection substantial change from the previous to the of source terms for MAGNEM is limited compared to the current Unified Rascal proposed Table R-1 values. The provided Interface (URI). A default mix from the FSAR was used. Fifteen-minute release calculations did not contain information that could duration was used.

be used to justify this change. Please provide justification that supports the changes in the Table All other parameters were duplicated for the Revision 6 Table R-1 values.

RA1.1, RS1.1, R-1 values from the previous values to the current 12 The NEI 99-01 Revision 6 Table R-1 calculation is documented in EPCI-14-02.

RG1.1 values.

URI software is the current dose assessment software used at the Callaway Energy Center. URI replaced MAGNEM to meet the requirement to calculate multiple release points. The URI source term of Clad Damage was selected based on this occurrence being more likely than a core melt down. The 1-hour release duration is based on NEI 99-01, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors." (For additional details, please refer to Attachment 4.)

13 N/A Omitted N/A Page 8 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For NEI 99-01, Revision 6, EALs AA3 and HA5, A review of all procedures associated with down power from 100% to Mode 5 CP is proposing two deviations. NEI Initiating (cold shutdown) reviewing field actions that have to be taken. After review, there Condition (IC) AA3 example 2, and HA5 example are no rooms which need to be accessed to shut the plant down to Mode 5.

1 will not be included because a review of CP RHR shutdown cooling does not have to be placed in service because CP can normal operating and shutdown procedures by cool down to Mode 5 using ASD, steam dumps and MSIV bypasses.

Operations Subject Matter Experts concluded that there are no areas external to the Main Control Closing the breakers for SI Accumulators and RHR loop suctions valves would Room that require access to perform a normal not be required.

plant shutdown and cooldown to Cold Shutdown The end result was that a shutdown to cold shutdown can be accomplished Generic conditions.

14 from the Control Room alone, without additional actions being performed.

AA3/HA5 a. Please verify that all required manipulations to The Control Room Ventilation System provides adequate protection from shut down the plant and enter shutdown cooling can be performed from the Main external hazardous gases.

Control Room or revise accordingly.

b. Please verify that no local breaker operations are required or revise accordingly.

c. Please provide evidence that an assessment of Control Room availability was performed to support these deviations.

Page 9 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL RU2.1, site-specific refueling pathway level indications are not provided per guidance in NEI 99-01, Revision 6. Additionally, the NEI 99-01 Basis discussion does not include the NEI 99-01, Revision 6, EAL AA2 guidance that This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.

RU2.1

a. Please provide site-specific level indications for RA2.1 EAL RU2.1 that could be used to support a. Added (EC LI-0039A, EC LI-0039B, local observation of SFP level)" as site-15 timely and accurate assessments; include specific SFP level indication to RU2.1.

RA2.2 applicable mode availability for this level RA2.3 instrumentation.

b. Please justify excluding the NEI 99-01, Revision 6, EAL AA2 guidance that relates to b. Added the cited applicability statement to the RA2.1 and RA2.2 bases:

RA2.1, RA2.2, and RA2.3 applicability or revise accordingly. This EAL applies to irradiated fuel that is licensed for dry storage up to the

c. Please verify that RA2.1 should be an Alert point that the loaded storage cask is sealed. Once sealed, damage to a and revise accordingly. loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1.1.

c. Corrected typo to read Alert.

Page 10 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL RA2.2, the logic was changed from NEI RA2.2 has been revised to read:

99-01, Revision 6, guidance that uses an increase Damage to irradiated fuel resulting in a release of radioactivity from the fuel as in radiation monitor readings to determine that indicated by any of the following:

irradiated fuel has been damaged to a proposed logic that requires the operator to know that

• Hi-Hi Alarm on Fuel Building exhaust monitors (GG-RE-27 or 28) 16 RA2.2 damage has occurred to irradiated fuel AND an there is an increase in radiation monitor

• Manipulator crane radiation monitor (SD-RE-41) >100 mR/hr indications. Please develop EAL RA2.2 per NEI

• Fuel Pool Bridge Crane OR Spent Fuel Pool Area radiation monitor 99-01, Revision 6, as endorsed or provide further (SD-RE-37 or 38) > 30 mR/hr justification for this deviation.

For EAL EU1.1, please explain why symbols were Revised EU1.1 by replacing the gamma and neutron symbols with the terms used rather than spelling out gamma and gamma and neutron.

neutron or revise accordingly.

Reworded the EAL as follows:

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading > EITHER of the following:

17 EU1.1

• 60 mrem/hr (gamma + neutron) on the top of the closure lid of the overpack

• 7,000 mrem/hr (gamma + neutron) on the side of the transfer cask This allows for this EAL to cover an accident while moving the spent fuel from the SFP to ISFSI. These numbers are 2x the TS limit. Refer to Certificate of Compliance No.1040, Appendix A, Section 5.3.4 (see Attachment 3).

Technical bases were revised to support this change.

For EAL CA1.1, a BBLI-53 A/B level of 0 inches is BBLI-53 A/B indications are trended and displayed when providing RCS level provided as an indication that RCS level is lower indication. Comparison between BBLI-53A and BBLI-53B, along with their than the bottom of the RCS hot leg. The Callaway trending together, provides adequate indication of failure of a channel.

Basis provides that BBLI-53A/B cannot sense (No document change) level changes in the Reactor Vessel below the 18 CA1.1 elevation of the RCS loop hot leg penetration.

Please provide justification that supports using the minimum value of BBLI-53 A/B for EAL classification as this reading may not be readily differentiated from an instrument failure or revise accordingly.

Page 11 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response 19 N/A Omitted N/A For EAL CS1.1 and CS1.2, the logic was changed Both CS1.1 and CS1.2 wording is consistent with the generic guidance intent.

from NEI 99-01, Revision 6, guidance without The status of containment closure modifies the level threshold consistent with CS1.1 justification. As changed, the EAL appears vague the generic guidance.

20 CS1.2 and interpretive. Please develop EAL CS1.1 and (No document change)

CS1.2 per NEI 99-01, Revision 6, as endorsed, or provide further justification for deviation.

For EALs CS1.3 and CG1.2, CP did not include Added of sufficient magnitude to indicate core uncovery to sump/tank level of sufficient magnitude to indicate core uncovery increases in CS1.3 and CG1.2.

to the unplanned increase in any sump/tank level to the IC wording. Additionally, CP added Visual CS1.3 observation of UNISOLABLE RCS leakage to the Deleted visual observation of unisolable RCS leakage from CS1.3 and CG1.2.

21 IC wording. As proposed, EALs CS1.3 and CG1.2 CG1.2 could result in unnecessary Site and General Emergency declarations. Please provide further justification or revise EAL CS1.3 and CG1.2 accordingly consistent with NEI 99-01, Revision 6, as endorsed.

For EAL CG1.2 and the Containment Fission Deleted reference to CA-3 in CG1.2.

Product Barrier Potential Loss D.2, it is not clear how CA-3, Hydrogen Flammability in Containment, can be used to estimate containment atmosphere hydrogen concentration.

22 CG1.2 Please explain how a procedure to determine hydrogen flammability in containment can be used to estimate containment atmosphere hydrogen concentration or remove the reference to CA-3 from Technical Basis.

Page 12 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EALs CU2.1, CA2.1, SU1.1, SA1.1, SS1.1, SG1.1, and SG1.2, AC power sources are provided by Table C-3. Additionally, the Callaway Basis provides that credit can be taken for additional sources of power if they are capable of carrying an emergency bus.

a. Omitted a. N/A

b. Please justify the addition of Additional b. Per the developer notes:

sources of offsite power are available from The EAL and/or Basis section may specify use of a non-safety-related diesel generators such as the Alternate power source provided that operation of this source is controlled in Emergency Power Supply (AEPS) or portable accordance with abnormal or emergency operating procedures, or beyond generation sources. Credit can be taken for design basis accident response guidelines (e.g., FLEX support guidelines).

these sources if they are capable of carrying Such power sources should generally meet the Alternate ac source an NB bus and are aligned within 15 minutes definition provided in 10 CFR 50.2 to the Callaway Basis as this statement could CU2.1, CA2.1, potentially be applied to power supplies not The Alternate Emergency Power Supply (AEPS) was built for, and has been 23 SA1.1, SS1.1, listed on Table C-3 or revise according to demonstrated to be capable of carrying one of the two Safety Related SG1.1, SG1.2 NRC endorsed guidance. Emergency Buses. Thus, it would be able to energize and maintain a Safety Related bus and provide the electric power needed to mitigate the consequences of an accident from this non-safety related power supply. In practice, without some initial setup it takes more than 15 minutes to line-up AEPS to a safety related bus. Thus the wording in Callaway's current EAL basis for SA1.1 stating "...if they are capable of carrying an NB bus and are aligned within 15 minutes".

The second half of this question concerning "or portable generation sources" has been put in place based on the future actions we intend to perform to comply with the FLEX requirements. This capability has not been put in place, but is anticipated that is will meet the requirements needed to energize a Safety Related 4.160 kV bus. It is not anticipated that Callaway will be able to meet the 15 minute requirement with the portable generator, but if already lined up it would maintain the bus energized and allow for mitigating the accident.

(No document change)

Page 13 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL CU3.1 and CA3.1, please explain how Deleted due to the loss of decay heat removal capability from CU3.1.

the addition of due to the loss of decay heat removal capability to EAL CU3.1 and due to a loss of RCS cooling to EAL CA3.1 would not Deleted due to a loss of RCS cooling from CA3.1. Added the following CU3.1 result in potential misclassification for an event omitted generic EAL wording:

24 CA3.1 other than a loss of decay heat removal that leads (This EAL does not apply during water-solid plant conditions.)

to an unplanned RCS temperature and/or RCS pressure rise. Please provide justification or revise accordingly consistent with endorsed guidance For EALs CU5.1 and SU7.1, the Sentry Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2.4.

Notification System is provided as an offsite "The Sentry System provides a means of performing required ORO emergency response organization (ORO) communication notifications in a timely manner (within 15 minutes of classification). "

method for the electronic transmission of a CU5.1 notification form to the OROs. Please provide When each ORO acknowledges receipt of a Sentry notification, Sentry 25 SU7.1 reference to specific section of the site emergency provides Callaway Plant with positive confirmation that ORO received the plan that that identifies the Sentry Notification message.

System as a means of timely notification to OROs for a spectrum of potential event responses or revise accordingly.

Page 14 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL CA3.1, Note 10: Begin monitoring hot Revised Note 10 to read:

condition EALs concurrently, was added to the Begin monitoring hot condition EALs concurrently for any new event or provided EAL Technical Basis. It is not clear to condition not related to the loss of decay heat removal.

the staff how Note 10 would be applied during an UNPLANNED increase in RCS temperature event. This note re-enforces the implementation guidance in NEI 99-01 that states:

Please provide justification for this difference or The mode in effect at the time that an event or condition occurred, and prior revise accordingly.

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 26 CA3.1 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.

The note reminds the end-user that hot condition EALs become applicable for any new event or condition once Mode 4 is entered from Mode 5 during a loss of decay heat removal.

(Note 10 revised throughout document and wallchart)

For EAL CA6.1 and SA9.1, the Callaway Basis Added the applicable portion of the seismic event discussion in HU2.1 to the discussion for seismic events refers to a CA6.1 and SA9.1 bases.

CA6.1 discussion under EAL HU2.1. Please include the 27 discussion on seismic events in the EAL CA6.1 SA9.1 and SA9.1 Callaway Basis or provide justification for not including the discussion as this could impact the timeliness of event assessment.

Page 15 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL HU2.1, the proposed EAL may not be The HU2.1 wording was intended for the Unusual Event classification to be consistent with from NEI 99-01, Revision 6, driven by receipt of seismic activity annunciator 98D (OBE) in the control room guidance, which provides that site-specific and not by actions performed in the seismic event AOP (OTO-SG-00001).

indication that a seismic event met or exceeded Revised HU2.1 to read OBE [operating basis earthquake] limits should be based on the indications, alarms, and displays Seismic event > OBE as indicated by Seismic Activity, Annunciator 98D.

of site-specific monitoring equipment. The 28 HU2.1 proposed EAL appears to base the declaration on implementation of an alarm response manual (OTO-SG-00001). Please provide justification for using OTO-SG-00001 for event classification rather than the appropriate seismic monitoring equipment as provided by NRC endorsed guidance or revise accordingly.

For EAL HU3.2, the proposed Callaway Basis identifies the Control Building, Battery Room, and ESF Switchgear Room as internal flooding areas of concern. Additionally, the Callaway Basis for HU3.2, which is applicable for all modes, references CA6.1, which is applicable in modes 5 and 6, for internal flooding affecting one or more safety trains.

29 HU3.2 a. Please explain how the statement in EAL HU3.2 that limits flooding areas of concern will a. Deleted bases statement related to internal flooding areas.

not potentially be used to limit a flooding related EAL declaration to only equipment in the Control Building, Battery Room, and the ESF Switchgear Room or revise accordingly.

b. Please explain why EAL HU3.2 only b. Added reference to SA9.1 in addition to CA6.1 in the HU3.2 bases.

references an EAL that is applicable in lower modes.

Page 16 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL HS5.1, please consider the following or provide an explanation how this EAL can be consistently applied:

st 1 bullet - Revised to restrict mode applicability of HS5.1 to Modes 1, 2, 3, 4,

• Addition of operating mode specificity to

5. None of the three listed safety functions are required when the reactor the listed safety functions to preclude vessel is defueled.

event classification when these safety functions are no longer needed in Restricted reactivity control safety function to Modes 1, 2 and 3 only since, by 30 HS5.1 accordance with site technical definition, the reactor has adequate shutdown margin while in Modes 4 and 5.

specifications; and

• Including a Clock start time in the nd 2 bullet - Added the following to the bases regarding clock start time:

Callaway Basis discussion.

For the purpose of this EAL the 15 minute clock starts when the last licensed operator leaves the Control Room.

For NEI 99-01, Revision 6, EAL SU3.1, CP does Callaway does not have a site-specific radiation monitor correlation that not provide an EAL that uses site-specific supports identifying a monitor reading that correspond to TS coolant activity radiation monitor(s). Please provide additional limits.

justification that a Callaway EAL cannot be The associated EAL is based on whether or not Technical Specification limits developed consistent with endorsed guidance or for RCS activity have been exceeded. Callaway's Technical Specifications revise accordingly.

include limits for Iodines governed by Dose Equivalent Iodine (DEI) and noble SU3 gases governed by Dose Equivalent Xenon (DEXe). DEI and DEXe both 31 represent weighted sums of measured isotopic concentrations. Determination

[SU4.1]

of DEI and DEXe is performed by laboratory analysis. Callaway's inline process radiation monitors do not have the capability to perform gamma spectrum measurements and determine concentrations of the specific isotopes that contribute to DEI or DEXe. Additionally, the process radiation monitors would not have the capability to perform weighted sums to compare measured count rates with applicable Technical Specification limits.

(No Document change)

Page 17 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EAL SU4.1, please explain why the proposed Revised SU4.1 to read:

wording is different from the NEI 99-01, Revision Sample analysis indicates RCS activity > Technical Specification Section 32 SU4.1 6, guidance which clearly states sample analysis 3.4.16 limits indicates that, or revise accordingly.

For EAL SU5.1, please explain how timely Callaway does not rely on a manual method of performing an RCS inventory declaration can be performed without reliance on balance. As stated in the bases:

a potentially time consuming manual method of Manual or computer-based methods of performing an RCS inventory performing an RCS inventory balance or revise balance are normally used to determine RCS leakage. The Personal the Callaway Basis accordingly.

Computer (PC) is preferred method of calculating RCS leak rate. When the PC is used, plant status information and all calculations are generated by the OSPBB9 software program. When the PC software is not available, 33 SU5.1 procedural guidance is available to perform the manual RCS inventory balance.

If the preferred method (computer) is not available, then the manual analysis method is performed, but only in the absence of computer based methods.

Both computer and manual methods can be completed within a 15 minute time period.

(No document change)

For EAL SU6.1, please provide a justification for The words subsequent automatic trip were added to both SU6.1 and SU6.2 to including a subsequent automatic trip to the EAL address the condition where an automatic trip signal other than the initial condition or revise accordingly automatic trip failure successfully shuts down the reactor prior to any manual trip action being initiated. For example, if the reactor receives a valid reactor 34 SU6.1 trip signal on high pressurizer pressure but fails to trip, but AMSAC automatically initiates and successfully trips the reactor before the manual trip signal was inserted, the EAL will still have been exceeded and an Unusual Event declared.

(No document change)

Page 18 of 19

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP)

DOCKET NO. 50-483 RAI-CP SECTION/EAL Question CP Response For EALs SU6.1, SU6.2, SA6.1, and SS6.1, The method used to determine that the reactor is shutdown following a reactor please provide further justification as to why trip, for the purposes of emergency classification, is consistent with the greater than or equal to five percent reactor power Callaway EOPs (E-0), i.e., indication of reactor power < 5%. This is also the was added or revise accordingly. (Note: power level that defines power operation in the Technical Specifications. As Westinghouse EOPs do not solely rely on Reactor specified in the generic developers guidance:

SU6.1, SU6.2, Power level to determine the status of reactor 35 Developers may include site-specific EOP criteria indicative of a successful SA6.1, SS6.1 criticality.)

reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level).

Reactor power < 5% is therefore the site-specific indication of a successful reactor trip for emergency classification.

(No document change)

Page 19 of 19

Attachment 2 to ULNRC-06227 Summary of EAL Changes NOT Associated with RAI Responses 1 Page

Summary of EAL Changes NOT Associated with RAI Responses The table below summarizes changes that have been introduced to the EAL submittal documentation for reasons other than the responses to the NRC RAIs.

Section/EAL Description 5.1 Definitions Added the abbreviation (OCA) to the Owner Controlled Area definition.

5.1 Definitions Added the abbreviation (PA) to the Protected Area definition.

5.2 ODCM - Removed the - from Off-site for consistency throughout the document.

Abbreviations/Acronyms 7.2 Corrected the Attachment 2 Title to Fission Product Barrier Loss / Potential Loss Matrix and Bases, by adding Loss / Potential Loss.

EU1.1 Corrected the section titled VCSNS Basis Reference to Callaway Basis Reference.

SU4.1 In the Basis, corrected XE-133 to Xe-133.

SU6.2 In the Basis, ninth paragraph, second sentence, corrected the sentence by adding trip after manually.

SU7.1 In the Basis, corrected the reference from Table C-5 to Table S-4.

SU7.1 In the Basis, added the sentence This EAL is the hot condition equivalent of the cold condition EAL CU5.1., to match the similar sentence in EAL CU5.1.

Fuel Clad In the Basis, deleted duplicate indicates in first sentence.

B.1.

Potential Loss RCS In the Category line, corrected B. CMT Radiation/RCS Activity to C. CMT Radiation/RCS Activity.

C Potential Loss RCS In the Basis section, corrected Emergency Director to Emergency Coordinator.

E.1 Potential Loss Page 1 of 1

Attachment 3 to ULNRC-06227 Basis for Spent Fuel Storage Cask Dose Rate Limits (Excerpt from Certificate of Compliance No 1040 Amendment No. 0, Appendix A) 1 Page

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued) 5.3 Radiation Protection Program 5.3.1 Each cask user shall ensure that the Part 50 radiation protection program appropriately addresses dry storage cask loading and unloading, as well as ISFSI operations, including transport of the loaded TRANSFER CASK outside of facilities governed by 10 CFR Part 50. The radiation protection program shall include appropriate controls for direct radiation and contamination, ensuring compliance with applicable regulations, and implementing actions to maintain personnel occupational exposures As Low As Reasonably Achievable (ALARA). The actions and criteria to be included in the program are provided below.

5.3.2 As part of its evaluation pursuant to 10 CFR 72.212(b)(2)(i)(C), the licensee shall perform an analysis to confirm that the dose limits of 10 CFR 72.104(a) will be satisfied under the actual site conditions and ISFSI configuration, considering the planned number of casks to be deployed and the cask contents.

5.3.3 Based on the analysis performed pursuant to Section 5.3.2, the licensee shall establish individual cask surface dose rate limits for the TRANSFER CASK and the VVM to be used at the site. Total (neutron plus gamma) dose rate limits shall be established at the following locations:

a. The top of the VVM.

b. The side of the TRANSFER CASK

c. The outlet vents on the VVM 5.3.4 Notwithstanding the limits established in Section 5.3.3, the average of the measured dose rates on a loaded VVM or TRANSFER CASK shall not exceed the following values:

a. 30 mrem/hr (gamma + neutron) on the top of the closure lid of the VVM

b. 3500 mrem/hr (gamma + neutron) on the side of the TRANSFER CASK 5.3.5 The licensee shall measure the TRANSFER CASK and VVM surface neutron and gamma dose rates as described in Section 5.3.8 for comparison against the limits established in Section 5.3.3 or Section 5.3.4, whichever are lower.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-3

Attachment 4 to ULNRC-06227 Basis for Difference in NEI 99-01 Revision 6 and Revision 5 Emergency Action Level Table R-1, "Effluent Monitor Classification Threshold Values" 1 Page

BASIS FOR DIFFERENCE IN NEI 99-01 REVISION 6 AND REVISION 5 EMERGENCY ACTION LEVEL TABLE R-1, "EFFLUENT MONITOR CLASSIFICATION THRESHOLD" VALUES The NEI 99-01 Revision 5 Table R-1 calculation is documented in EPCI-08-01.

The software program MAGNEM was used.

Selection of source terms for MAGNEM is limited compared to the Unified Rascal Interface (URI). A default mix from the FSAR was used.

Fifteen minute release duration was used.

All other parameters were duplicated for the Revision 6 Table R-1 values.

REVISON 5 REVISION 6 Calculation Software MAGNEM Unified RASCAL Interface Release Duration 15 Minutes 1 Hour Source Term Default mix (FSAR) Clad Damage The NEI 99-01 Revision 6 Table R-1 calculation is documented in EPCI-14-02.

URI software is the current dose assessment software used at the Callaway Energy Center.

URI replaced MAGNEM to meet the requirement to calculate multiple release points.

The URI source term of Clad Damage was selected based on this occurrence being more likely than a core melt down.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> release duration is based on NEI 99-01, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors.

Table One compares the different calculational values based on the change in release duration and using core melt or clad damage.

TABLE ONE Revision UV GE ASD GE AFW GE Rev 5, MAGNEM, Release Duration 0.25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (Current Table R-1 Values EPCI-08-91) 5.28 E+8 358 2,080 Rev 5, MAGNEM, Release Duration 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.46 E+8 89.5 520 Rev 6, URI Release Duration 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, Core Clad damage (Values used in Revision 6 EPCI-14-02) 6.59 E+7 12 163 Rev 6, URI Release Duration 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, Core Melt 1.61 E+8 29 391 Page 1 of 1

Attachment 5 to ULNRC-06227 Calculation EPCI-08-01 Rev. 1, "Dose Projection Calculations to Support NEI 99-01 Rev.5 Emergency Action Levels" 7 Pages

EPCI- _08_- _01_ Rev. 001 CALLAWAY PLANT EMERGENCY PREPAREDNESS CALCULATION COVER SHEET Page ~1 __ of_7_

TITLE: Dose Projection Calculations to Support NEI 99-01 Rev.S Emergency Action Levels Purpose and Scope These calculations are being performed to provide a basis for threshold values in Table R-1 of the Emergency Action Levels. The calculations are performed using the MAGNEM Version 21.1.2 dose projection software.

Assumptions and References The following assumptions and constants were used to perform the calculations:

1. The software uses equations/methodology as described in EPCI 94-03, MAGNEM Dose Calculation.

2. Source Term is the default mix per FSAR and EPCI 94-03.

3. Stability Class is "D" for all calculations based on the RERP and Table 2.3-36 of the FSAR.

4. Wind Speed is 3.5 meters/second (7.8 mph) based on Table 2.3-17 of the FSAR.

5. Calculations are based on a 15 minute release at the radiation monitor threshold value, to reach the applicable trigger point (i.e., EPA Protective Action Guides). To accomplish this, since the software only allows release duration to be set in tenths of hours, duration and trigger values were multiplied by four.

6. Calculations are assumed to be performed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reactor shutdown and release start.

7. Wind Direction is 180 degrees, but does not affect the calculations.

8. The trigger points are based on:

a. Site Emergency= 100 mrem TEDE/500 mrem CDE Thyroid

b. General Emergency= 1 Rem TEDE/5 Rem CDE Thyroid PREPARED BY: ~ 7914/Eme<g. Re.ponse Coord./8-20*08 (Signature/PIN/Title/Date) r<

INDEPENDENTLY/7(\nt(

• t REVIEWEDBY: ~ 1 f '3~D4 7f.~-n\~C\~f 'f/'Z 06 j,_)\T:~

(Signat e/PIN/Title/Date)

APPROVED BY:

File K191.0018 Page 1 of 1 CA2264 02/19/08 KDP-ZZ-00007

EfC:L D~-o/

MAGNEM 21.1 Printed Output Report

,{'ef. tJOI Calculation ID: 8877758 z_ IS[' 7 Input Source:

Input File:

Plant

\\calntpfile2\magnemdata\MAGNEM.DAT ~ f-w~c'f Input Date/Time: 08:59 08/20/2008 ncnt.y Operator Location: EOF Accident Type: LOCA Reactor Shutdown: YES Reactor Shutdown Time: 07:00 Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: General Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: Unit Vent ( GT-RE-21B Reading: 5.82E+08 uCi/sec Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 (Rem)

EAB TEDE: 1.381 2 Mile TEDE: 0.365 5 Mile TEDE: 0.100 10 Mile TEDE: 0.037 EAB Thyroid:

2 Mile Thyroid:

~ 5.284 5 Mile Thyroid: 1. 453 10 Mile Thyroid: 0.542 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: EVACUATE 2 Mile: EVACUATE 5 Mile: EVACUATE 10 Mile: NONE PAG based on Thyroid exceeded at 2.0 miles

============================

End of output report

MAGNEM 21.1 Printed Output Report Elc_r 6f-o;

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • f(..R_l. (90 I Calculation ID: 4421693 Input Source: Plant 3(>f 7 Input File:

Input Date/Time:

\\calntpfile2\magnemdata\MAGNEM.DAT 09:03 08/20/2008 j/t f-  ;?D -b$

Operator Location: EOF Accident Type: SGTR (Direct to Atmos.)

Reactor Shutdown: YES Reactor Shutdown Time: 07:00 Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: General Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: "A" ASD ( AB-RE-111 Reading: 3.58E+02 mr/hr Flowrate: 5.95E+05 lb/hr Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 (Rem)

EAB 2 Mile TEDE:

TEDE:

~1.054 5 Mile TEDE: 0.288 10 Mile TEDE: 0.107 EAB Thyroid: 14.131 2 Mile Thyroid: 3.724 5 Mile Thyroid: 1.019 10 Mile Thyroid: 0.377 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: EVACUATE 2 Mile: EVACUATE 5 Mile: EVACUATE 10 Mile: NONE PAG based on TEDE exceeded at 2.0 miles

============================

End of output report

MAGNEM 21.1 Printed Output Report Calculation ID: 8145593 Input Source: Plant Input File: \\calntpfile2\magnemdata\MAGNEM.DAT Input Date/Time: 09:06 08/20/2008 Operator Location: EOF Accident Type: SGTR (Direct to Atmos.)

Reactor Shutdown: YES Reactor Shutdown Time: 07:00 CW- q{7-/of)

Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: General Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: AFTD ( FC-RE-385 Reading: 2.08E+03 mr/hr Flowrate: 7.35E+04 lb/hr Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 (Rem)

~

EAB TEDE:

2 Mile TEDE: 4 5 Mile TEDE: 0.288 10 Mile TEDE: 0.107 EAB Thyroid: 14.125 2 Mile Thyroid: 3.723 5 Mile Thyroid: 1.019 10 Mile Thyroid: 0.377 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: EVACUATE 2 Mile: EVACUATE 5 Mile: EVACUATE 10 Mile: NONE PAG based on TEDE exceeded at 2.0 miles

============================

End of output report

MAGNEM 21.1 Printed Output Report &fc21 of-DI

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • Ke,J 00 I Calculation ID: 6525427 Input Source: Plant ~of7 Input File:

Input Date/Time:

\\calntpfile2\magnemdata\MAGNEM.DAT 09:01 08/20/2008 1/1 f-Zll --o(

Operator Location: EOF Accident Type: LOCA Reactor Shutdown: YES Reactor Shutdown Time: 07:00 Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: Site Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: Unit Vent ( GT-RE-21B Reading: 5.82E+07 uCi/sec Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 (Rem)

EAB TEDE: 0.138 2 Mile TEDE: 0.036 5 Mile TEDE: 0.010 10 Mile TEDE: 0.004

~

EAB Thyroid:

2 Mile Thyroid: 8 5 Mile Thyroid: 0.145 10 Mile Thyroid: 0.054 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: NONE 2 Mile: NONE 5 Mile: NONE 10 Mile: NONE PAG based on Thyroid exceeded at 0.4 miles

============================

End of output report

MAGNEM 21.1 Printed Output Report £/e__r Df-D/

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ~e.V. OD(

Calculation ID: 8519082 Input Source: Plant 0 Df 7 Input File: \\calntpfile2\magnemdata\MAGNEM.DAT Input Date/Time: 09:03 08/20/2008 1/-zo-of Operator Location: EOF Accident Type: SGTR (Direct to Atmos.)

Reactor Shutdown: YES Reactor Shutdown Time: 07:00 Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: Site Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: "A" ASD ( AB-RE-111 Reading: 3.59E+01 mr/hr Flowrate: 5.95E+05 lb/hr Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 EAB TEDE:

2 Mile TEDE:

5 Mile TEDE:

~ ~

0.029 6

10 Mile TEDE: 0 011 0

EAB Thyroid: 1.417 2 Mile Thyroid: 0.373 5 Mile Thyroid: 0.102 10 Mile Thyroid: 0.038 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: NONE 2 Mile: NONE 5 Mile: NONE 10 Mile: NONE PAG based on TEDE exceeded at 0.4 miles End of output report

MAGNEM 21.1 Printed Output Report £/cr o?~o;

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ;?e- ,; oo r Calculation ID:

Input Source:

2534861 Plant 7 Df 7 Input File: \\calntpfile2\magnemdata\MAGNEM.DAT Input Date/Time: 09:06 08/20/2008

/It P-~v- Df' Operator Location: EOF Accident Type: SGTR (Direct to Atmos.)

Reactor Shutdown: YES Reactor Shutdown Time: 07:00 Reactor Shutdown Date: 08/20/2008 Release Start Time: 07:00 Release Start Date: 08/20/2008 Accident Classification: Site Emergency Release Duration: 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Wind Speed: 7.8 MPH Wind Direction from: 180.0 Stability Class: D Isotopic Mix: Default Mix

• • • • • EFFLUENT MONITOR DATA *****

Monitor: AFTD ( FC-RE-385 Reading: 2.09E+02 mr/hr Flowrate: 7.35E+04 lb/hr Calculation Time: 08:00 Calculation Date: 08/20/2008 PROJECTED DOSES -- calculated at 08:00 08/20/2008 EAB TEDE:

2 Mile TEDE:

d:b .

0.106 5 Mile TEDE: 0.029 10 Mile TEDE: 0. Oll EAB Thyroid: 1.417 2 Mile Thyroid: 0.373 5 Mile Thyroid: 0.102 10 Mile Thyroid: 0.038 Affected Sectors EAB: ALL 2 Mile: ALL 5 Mile: R,A,B 10 Mile: R,A,B PARs EAB: NONE 2 Mile: NONE 5 Mile: NONE 10 Mile: NONE PAG based on TEDE exceeded at 0.4 miles

============================

End of output report

Attachment 6 to ULNRC-06227 Callaway NEI 99-01 Revision 6 EAL Wall Charts (Information Only) 3 Pages

Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE dose greater than 10 mrem TEDE or 50 mrem thyroid CDE the ODCM limits for 60 minutes or longer RG1.1 1 2 3 4 5 6 DEF RS1.1 1 2 3 4 5 6 DEF RA1.1 1 2 3 4 5 6 DEF RU1.1 1 2 3 4 5 6 DEF Reading on any Table R-1 effluent radiation monitor Reading on any Table R-1 effluent radiation monitor Reading on any Table R-1 effluent radiation monitor Reading on any Table R-1 effluent radiation monitor

> column "GE" for 15 min. (Notes 1, 2, 3, 4) > column "SAE" for 15 min. (Notes 1, 2, 3, 4) > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) > column "UE" for 60 min. (Notes 1, 2, 3)

RA1.2 1 2 3 4 5 6 DEF RU1.2 1 2 3 4 5 6 DEF RG1.2 1 2 3 4 5 6 DEF RS1.2 1 2 3 4 5 6 DEF Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or Sample analyses for a gaseous or liquid release indicates Dose assessment using actual meteorology indicates Dose assessment using actual meteorology indicates a concentration or release rate 2 x ODCM limits for 60 beyond the SITE BOUNDARY (Note 4) doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at doses > 100 mrem TEDE or 500 mrem thyroid CDE at or min. (Notes 1, 2) or beyond the SITE BOUNDARY (Note 4) beyond the SITE BOUNDARY (Note 4) RA1.3 1 2 3 4 5 6 DEF Rad Analysis of a liquid effluent sample indicates a Effluent RG1.3 1 2 3 4 5 6 DEF RS1.3 1 2 3 4 5 6 DEF concentration or release rate that would result in doses >

10 mrem TEDE or 50 mrem thyroid CDE at or beyond the Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) beyond the SITE BOUNDARY: beyond the SITE BOUNDARY:

Closed window dose rates > 100 mR/hr expected to RA1.4 1 2 3 4 5 6 DEF Closed window dose rates > 1000 mR/hr expected continue for 60 min. Field survey results indicate EITHER of the following at or to continue for 60 min.

Analyses of field survey samples indicate thyroid beyond the SITE BOUNDARY:

Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. Closed window dose rates > 10 mR/hr expected to CDE > 5000 mrem for 60 min. of inhalation.

(Notes 1, 2) continue for 60 min.

(Notes 1, 2)

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

(Notes 1, 2)

Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, Unplanned loss of water level above irradiated fuel fuel racks for 60 minutes or longer irradiated fuel RG2.1 1 2 3 4 5 6 DEF RS2.1 1 2 3 4 5 6 DEF RA2.1 1 2 3 4 5 6 DEF RU2.1 1 2 3 4 5 6 DEF Abnormal Rad Spent fuel pool level cannot be restored to at least 2022 ft. Lowering of spent fuel pool level to 2022 ft. 1.25 in. (Level 3) Uncovery of irradiated fuel in the REFUELING PATHWAY UNPLANNED water level drop in the REFUELING PATHWAY Levels 1.25 in. (Level 3) for 60 min. (Note 1) as indicated by low water level alarm or indication

/ (EC LI-0039A)

RA2.2 1 2 3 4 5 6 DEF Table R-1 Effluent Monitor Classification Thresholds AND Rad Damage to irradiated fuel resulting in a release of Effluent UNPLANNED rise in corresponding area radiation levels as Release Point Monitor GE SAE Alert UE radioactivity from the fuel as indicated by any of the indicated by any Table R-2 radiation monitors following:

Unit Vent GT-RE-21B 6.59E+7 µCi/sec 6.59E+6 µCi/sec 6.59E+5 µCi/sec 2 X Hi-Hi alarm Irradiated Hi-Hi Alarm on Fuel Building exhaust monitors Fuel Event (GG-RE-27 or 28)

AB-RE-111/

Gaseous ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr ----- ----- Manipulator crane radiation monitor (SD-RE-41) 112/113/114

>100 mR/hr TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ----- Fuel Pool Bridge Crane OR Spent Fuel Pool Area radiation monitor (SD-RE-37 or 38) > 30 mR/hr Radwaste Bldg Vent GH-RE-10B ----- ----- ----- 2 X Hi-Hi alarm RA2.3 1 2 3 4 5 6 DEF Liquid Liquid Radwaste Lowering of spent fuel pool level to 2031 ft. 1.25 in. (Level 2)

HB-RE-18 ----- ----- ----- 2 X Hi-Hi alarm Discharge Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown Table R-2 Fuel Building & Containment Area Radiation Monitors RA3.1 1 2 3 4 5 6 DEF Dose rate > 15 mR/hr in EITHER of the following areas:

None Fuel Building: Control Room (SD-RE-33)

Area SD-RE-34, Cask Handle Area Radiation Central Alarm Station (by survey)

Radiation SD-RE-35, New Fuel Storage Area Radiation Levels SD-RE-36, New Fuel Storage Area Radiation SD-RE-37, Fuel Pool Bridge Crane Radiation SD-RE-38, Spent Fuel Pool Area Radiation Containment:

SD-RE-40, Personnel Access Hatch Area Damage to a loaded cask CONFINEMENT BOUNDARY SD-RE-41, Manipulator Crane Radiation Monitor SD-RE-42, Containment Building Radiation EU1.1 1 2 3 4 5 6 DEF GT-RE-59 Containment High Area Radiation Monitor Damage to a loaded cask CONFINEMENT BOUNDARY as None GT-RE-60 Containment High Area Radiation Monitor None indicated by an on-contact radiation reading > EITHER of the Confinement following:

ISFSI Boundary 60 mrem/hr (gamma + neutron) on the top of the closure lid of the overpack 7,000 mrem/hr (gamma + neutron) on the side of the transfer cask HOSTILE ACTION resulting in loss of physical control of the HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or Confirmed SECURITY CONDITION or threat facility airborne attack threat within 30 minutes HG1.1 1 2 3 4 5 6 DEF HS1.1 1 2 3 4 5 6 DEF HA1.1 1 2 3 4 5 6 DEF HU1.1 1 2 3 4 5 6 DEF A HOSTILE ACTION is occurring or has occurred within A HOSTILE ACTION is occurring or has occurred within A HOSTILE ACTION is occurring or has occurred within A SECURITY CONDITION that does not involve a the PROTECTED AREA as reported by the Security Shift the PROTECTED AREA as reported by the Security the OWNER CONTROLLED AREA as reported by the HOSTILE ACTION as reported by Security Shift Supervisor Shift Supervisor Security Shift Supervisor Supervisor AND EITHER of the following has occurred: OR OR Security Any of the following safety functions cannot be A validated notification from NRC of an aircraft attack Notification of a credible security threat directed at the site controlled or maintained threat within 30 min. of the site OR

- Reactivity control A validated notification from the NRC providing information

- Core cooling of an aircraft threat

- RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Seismic event greater than OBE level None None None HU2.1 1 2 3 4 5 6 DEF Seismic Seismic event > OBE as indicated by Seismic Activity, Event Annunciator 98D Hazardous event HU3.1 1 2 3 4 5 6 DEF Note 1: The Emergency Coordinator should declare the A tornado strike within the PROTECTED AREA event promptly upon determining that time limit has been exceeded, or will likely be exceeded HU3.2 1 2 3 4 5 6 DEF Note 2: If an ongoing release is detected and the release Internal room or area FLOODING of a magnitude start time is unknown, assume that the release sufficient to require manual or automatic electrical duration has exceeded the specified time limit isolation of a SAFETY SYSTEM component needed for the current operating mode None None Natural or Note 3: If the effluent flow past an effluent monitor is Tech. known to have stopped, indicating that the HU3.3 1 2 3 4 5 6 DEF Hazard release path is isolated, the effluent monitor reading is no longer VALID for classification Movement of personnel within the PROTECTED AREA purposes is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or Note 4: The pre-calculated effluent monitor values toxic gas release) presented in EALs RA1.1, RS1.1 and RG1.1 HU3.4 1 2 3 4 5 6 DEF should be used for emergency classification assessments until the results from a dose A hazardous event that results in on-site conditions assessment using actual meteorology are sufficient to prohibit the plant staff from accessing the available site via personal vehicles (Note 7)

Note 5: If the equipment in the listed room or area was FIRE potentially degrading the level of safety of the plant already inoperable or out-of-service before the event occurred, then no emergency classification is warranted HU4.1 1 2 3 4 5 6 DEF A FIRE is not extinguished within 15 min. of any of the Note 6: If CONTAINMENT CLOSURE is re-established Table H-1 Fire Areas following FIRE detection indications (Note 1):

prior to exceeding the 30-minute time limit, Report from the field (i.e., visual observation) declaration of a General Emergency is not Area 5 Receipt of multiple (more than 1) fire alarms or required Containment indications Aux Feed Pump Rooms Field verification of a single fire alarm Note 7: This EAL does not apply to routine traffic Auxiliary Building AND impediments such as fog, snow, ice, or vehicle Diesel Generator Building The FIRE is located within any Table H-1 area breakdowns or accidents UHS Cooling Tower Hazards Note 8: A manual trip action is any operator action, or ESW Pumphouse HU4.2 1 2 3 4 5 6 DEF set of actions, which causes the control rods to Control Building/

be rapidly inserted into the core, and does not Communications Corridor Receipt of a single fire alarm (i.e., no other indications of include manually driving in control rods or RWST a FIRE) implementation of boron injection strategies Fuel Building AND None The fire alarm is indicating a FIRE within any Table H-1 Note 9: One Containment Spray System train and one area Containment Cooling System train comprise AND Fire one full train of depressurization equipment The existence of a FIRE is not verified within 30 min. of Note 10: Begin monitoring hot condition EALs concurrently alarm receipt (Note 1) for any new event or condition not related to the loss of decay heat removal HU4.3 1 2 3 4 5 6 DEF A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

HU4.4 1 2 3 4 5 6 DEF A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to Room alternate locations HS5.1 1 2 3 4 5 6 HA5.1 1 2 3 4 5 6 DEF An event has resulted in plant control being transferred from An event has resulted in plant control being transferred the Control Room to the Auxiliary Shutdown Panel (ASP) from the Control Room to the Auxiliary Shutdown Panel None Control AND (ASP) None Room Control of any of the following key safety functions is not Evacuation reestablished within 15 min. (Note 1):

Reactivity control (Mode 1, 2 and 3 only)

Core cooling RCS heat removal Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of General Emergency Coordinator warrant declaration of Site Area Emergency Coordinator warrant declaration of an Alert Coordinator warrant declaration of a UE HG6.1 1 2 3 4 5 6 DEF HS6.1 1 2 3 4 5 6 DEF HA6.1 1 2 3 4 5 6 DEF HU6.1 1 2 3 4 5 6 DEF Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress Emergency Coordinator indicate that events are in progress Emergency Coordinator, indicate that events are in progress Emergency Coordinator indicate that events are in progress or or have occurred which involve actual or IMMINENT or have occurred which involve actual or likely major failures or have occurred which involve an actual or potential have occurred which indicate a potential degradation of the substantial core degradation or melting with potential for of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or level of safety of the plant or indicate a security threat to loss of containment integrity or HOSTILE ACTION that HOSTILE ACTION that results in intentional damage or a security event that involves probable life threatening risk facility protection has been initiated. No releases of results in an actual loss of physical control of the facility. malicious acts, (1) toward site personnel or equipment that to site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring Releases can be reasonably expected to exceed EPA could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Judgment access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs Protective Action Guideline exposure levels offsite for more than the immediate site area. Any releases are not expected to result in exposure levels exposure levels.

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

EIP-ZZ-00101, Addendum 1 Rev.[xx]

EAL Classification Matrix Page 1 of 3 Power Operation Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled Prepared for Ameren by: Operations Support Services, Inc. - 6/11/15

Prolonged loss of all offsite and all onsite AC power to Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses Loss of all offsite AC power capability to emergency buses emergency buses for 15 minutes or longer for 15 minutes or longer for 15 minutes or longer SG1.1 1 2 3 4 SS1.1 1 2 3 4 SA1.1 1 2 3 4 SU1.1 1 2 3 4 Loss of all offsite and all onsite AC power capability, Table Loss of all offsite and all onsite AC power capability, Table AC power capability, Table S-1, to emergency 4.16KV Loss of all offsite AC power capability, Table S-1, to S-1, to emergency 4.16KV buses NB01 and NB02 S-1, to emergency 4.16KV buses NB01 and NB02 for 15 buses NB01 and NB02 reduced to a single power source emergency 4.16KV buses NB01 and NB02 for 15 min.

min. (Note 1) for 15 min. (Note 1) (Note 1)

AND EITHER:

AND Restoration of at least one emergency bus in < 4 Any additional single power source failure will result in Table S-1 AC Power Supplies hours is not likely (Note 1)

Loss of loss of all AC power to SAFETY SYSTEMS CSFST Core Cooling-RED Path conditions met Emergency Offsite:

AC Power Safeguards XFMR A or B via ESF LTC XFMR XNB01 Loss of all AC and vital DC power sources for 15 minutes or Startup XFMR XMR01 via ESF LTC XFMR XNB02 longer Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned)

SG1.2 1 2 3 4 Onsite:

EDG NE01 Loss of all offsite and all onsite AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 for 15 min. EDG NE02 AND Loss of all 125 VDC power based on battery bus voltage Loss of all vital DC power for 15 minutes or longer indications < 107 VDC on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for 15 min. (Note 1) SS2.1 1 2 3 4 None None Loss of Loss of all 125 VDC power based on battery bus voltage Vital DC indications < 107 VDC on all vital DC buses NK01, NK03 Power (Division 1) and NK02, NK04 (Division 2) for 15 min. (Note 1)

UNPLANNED loss of Control Room indications for 15 minutes or UNPLANNED loss of Control Room indications for 15 minutes or Table S-2 Safety System Parameters longer with a significant transient in progress longer SA3.1 1 2 3 4 SU3.1 1 2 3 4 Reactor power Loss of None RCS level An UNPLANNED event results in the inability to monitor one An UNPLANNED event results in the inability to monitor Control or more Table S-2 parameters from within the Control Room one or more Table S-2 parameters from within the RCS pressure Room for 15 min. (Note 1) Control Room for 15 min. (Note 1)

Indications Core Exit T/C temperature AND Level in at least one S/G Any significant transient is in progress, Table S-3 Auxiliary or emergency feed flow in at least one S/G Reactor coolant activity greater than Technical Specification allowable limits None None Table S-3 Significant Transients SU4.1 1 2 3 4 RCS Activity Sample analysis indicates RCS activity > Technical Specification Section 3.4.16 limits Reactor trip Runback 25% thermal power RCS leakage for 15 minutes or longer Electrical load rejection > 25%

electrical load SU5.1 1 2 3 4 None ECCS actuation RCS unidentified or pressure boundary leakage > 10 gpm for 15 min.

None None OR None RCS RCS identified leakage > 25 gpm for 15 min.

Leakage OR Leakage from the RCS to a location outside containment

> 25 gpm for 15 min.

(Note 1)

System Malfunct.

Inability to shut down the reactor causing a challenge to core Automatic or manual trip fails to shut down the reactor and Automatic or manual trip fails to shut down the reactor cooling or RCS heat removal subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor SS6.1 1 SA6.1 1 SU6.1 1 An automatic or manual trip fails to shut down the reactor An automatic or manual trip fails to shut down the reactor as An automatic trip did not shut down the reactor as as indicated by reactor power 5% indicated by reactor power 5% indicated by reactor power 5% after any RTS setpoint AND is exceeded AND All actions to shut down the reactor are not successful as AND indicated by reactor power 5% Manual trip actions taken at the reactor control console None A subsequent automatic trip or manual trip action taken AND EITHER: (SB-HS-1 or SB-HS-42) are not successful in shutting down at the reactor control consoles (SB-HS-1 or SB-HS-42)

CSFST Core Cooling-RED Path conditions met the reactor as indicated by reactor power 5% (Note 8) is successful in shutting down the reactor as indicated RTS CSFST Heat Sink-RED Path conditions met by reactor power < 5% (Note 8)

Failure SU6.2 1 A manual trip did not shut down the reactor as indicated by reactor power 5% after any manual trip action was initiated AND Table S-4 Communications Methods A subsequent automatic trip or manual trip action taken at the reactor control console (SB-HS-1 or SB-HS-42) is successful in shutting down the reactor as indicated by System Onsite ORO NRC reactor power < 5% (Note 8)

Gaitronics X Loss of all onsite or offsite communications capabilities Plant Radios X SU7.1 1 2 3 4 Plant Emergency Dedicated Phones X Plant Telephone System X X X Loss of all Table S-4 onsite communication methods None ENS (Red Phone) Line X X OR Loss of Loss of all Table S-4 ORO communication methods Comm. Note 1: The Emergency Coordinator should declare the Back-Up Radio System X OR event promptly upon determining that time limit Sentry Notification System X Loss of all Table S-4 NRC communication methods has been exceeded, or will likely be exceeded Note 8: A manual action isNone any operator action, or set of actions, which causes the control rods to be Failure to isolate containment or loss of containment pressure rapidly inserted into the core, and does not control include manually driving in control rods or SU8.1 1 2 3 4 implementation of boron injection strategies Note 9: One Containment Spray System train and one Any penetration is not isolated within 15 min. of a VALID Containment Cooling System train comprise containment isolation signal CMT one full train of depressurization equipment None None OR Isolation Containment pressure > 27 psig with < one full train of Failure containment depressurization equipment operating per design for 15 min. (Note 9)

(Note 1)

Hazardous event affecting a SAFETY SYSTEM Table S-5 Hazardous Events needed for the current operating mode SA9.1 1 2 3 4 Seismic event (earthquake)

The occurrence of any Table S-5 hazardous event Internal or external FLOODING event Hazardous High winds or tornado strike AND EITHER: None Event None Event damage has caused indications of degraded FIRE Affecting performance in at least one train of a SAFETY Safety EXPLOSION SYSTEM needed for the current operating mode Systems Other events with similar hazard The event has caused VISIBLE DAMAGE to a characteristics as determined by the SAFETY SYSTEM component or structure needed Emergency Coordinator for the current operating mode FG1.1 1 2 3 4 FS1.1 1 2 3 4 FA1.1 1 2 3 4 Loss of any two barriers Loss or potential loss of any two barriers (Table F-1) Any loss or any potential loss of either Fuel Clad or RCS None Fission Product AND (Table F-1)

Barrier Degradation Loss or potential loss of third barrier (Table F-1)

1. An automatic or manual ECCS 1. Operation of a standby charging 1. A leaking or RUPTURED SG is (SI) actuation required by pump is required by EITHER: FAULTED outside of containment EITHER: UNISOLABLE RCS leakage RCS or SG Tube None None None UNISOLABLE RCS leakage SG tube leakage Leakage SG tube RUPTURE

2. CSFST Integrity-RED Path conditions met

1. CSFST Core Cooling-RED 1. CSFST Core Cooling-ORANGE Path 1. CSFST Heat Sink-RED Path 1. CSFST Core Cooling-RED Path Path conditions met conditions met conditions met conditions met AND AND

2. CSFST Heat Sink-RED Path None Heat sink required None Restoration procedures not Inadequate Heat conditions met effective within 15 min. (Note1)

Removal AND Heat sink required

1. Containment radiation > 2.80E+03 1. Containment radiation 1. Containment radiation >

R/hr on GT-RE-59 (591) or > 6.40E+00 R/hr on GT-RE-59 (591) 8.06E+04 R/hr on GT-RE-59 GT-RE-60 (601) or GT-RE-60 (601) (591) or GT-RE-60 (601

2. Dose equivalent I-131 coolant None None None CMT Radiation / activity > 300 µCi/cc RCS Activity 3. CVCS letdown radiation

> 2.50E+01 µCi/ml on SJ-RE-01 (016)

1. Containment isolation is required 1. CSFST Containment-RED Path AND EITHER: conditions met Containment integrity has been 2. Containment hydrogen lost based on Emergency concentration 4%

Coordinator judgment 3. Containment pressure > 27 psig None None None None CMT Integrity or UNISOLABLE pathway from with < one full train of Bypass containment to the environment Containment depressurization exists equipment operating per design for 15 min. (Note 1, 9)

2. Indications of RCS leakage outside of containment

1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the Emergency Coordinator that Emergency Coordinator that indicates Emergency Coordinator that Emergency Coordinator that indicates Emergency Coordinator that Emergency Coordinator that Judgment indicates loss of the Fuel Clad potential loss of the Fuel Clad barrier indicates loss of the RCS barrier potential loss of the RCS barrier indicates loss of the Containment indicates potential loss of the barrier barrier Containment barrier EIP-ZZ-00101, Addendum 1, Rev.[xx]

EAL Classification Matrix Page 2 of 3 Power Operation Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled Prepared for Ameren by: Operations Support Services, Inc. - 6/11/15

Loss of RCS inventory affecting fuel clad integrity with Loss of RCS inventory affecting core decay heat removal Loss of RCS inventory UNPLANNED loss of RCS inventory for 15 minutes or longer Containment challenged capability CG1.1 5 6 CS1.1 5 6 CA1.1 5 6 CU1.1 5 6 RVLIS Pumps Off < 65% (Top of Fuel) for > 30 min. (Note 1) With CONTAINMENT CLOSURE not established, RVLIS Loss of RCS inventory as indicated by Reactor Vessel level UNPLANNED loss of reactor coolant results in RCS level less AND Pumps Off < 72% < bottom of RCS hot leg ID (RVLIS Pumps Off < 73% or than a required lower limit for 15 min.

Any Containment Challenge indication, Table C-2 BBLI-53 A/B at 0 inches) (Note 1)

CS1.2 5 6 CG1.2 5 6 With CONTAINMENT CLOSURE established, RVLIS CA1.2 5 6 CU1.2 5 6 RCS level cannot be monitored for 30 min. (Note 1) Pumps Off < 65% (Top of Fuel)

RCS water level cannot be monitored for 15 min. (Note 1) RCS water level cannot be monitored AND 5 6 AND EITHER AND EITHER:

CS1.3 Core uncovery is indicated by any of the following: UNPLANNED increase in any Table C-1 Sump / Tank UNPLANNED increase in any Table C-1 sump/

RCS water level cannot be monitored for 30 min. (Note 1) level tank level due to a loss of RCS inventory UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude to indicate core uncovery AND Visual observation of UNISOLABLE RCS leakage Visual observation of UNISOLABLE RCS leakage RCS Level Manipulator crane radiation monitor SD-RE-41 Core uncovery is indicated by any of the following:

> 10,000 mR/hr UNPLANNED increase in any Table C-1 sump/tank Erratic Source Range Monitor indication level of sufficient magnitude to indicate core uncovery Table C-1 Sumps/Tanks AND Manipulator crane radiation monitor SD-RE-41 Any Containment Challenge indication, Table C-2 > 10,000 mR/hr Containment Sumps Erratic Source Range Monitor indication Containment Normal Sumps Containment Instrument Sump Table C-2 Containment Challenge Indications PRT RCDT CONTAINMENT CLOSURE not established (Note 6) Auxiliary Building Sump Containment hydrogen concentration 4%

Unplanned rise in Containment pressure Table C-3 AC Power Supplies Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses for 15 for greater than 15 minutes minutes or longer Offsite:

Safeguards XMFR A or B via ESF LTC XMFR XNB01 CA2.1 5 6 DEF CU2.1 5 6 DEF Startup XMFR XMR01 via ESF LTC XMFR XNB02 Loss of all offsite and all onsite AC power capability, Table AC power capability, Table C-3, to emergency 4.16KV None Loss of Main XMFR XMA01 backfed via UAT XMFR XMA02 C-3, to emergency 4.16KV buses NB01 and NB02 for buses NB01 and NB02 reduced to a single power source for Emergency (only if already aligned) 15 min. (Note 1) 15 min. (Note 1)

AC Power Onsite: AND EDG NE01 Any additional single power source failure will result in loss EDG NE02 of all AC power to SAFETY SYSTEMS Inability to maintain plant in cold shutdown UNPLANNED increase in RCS temperature Table C-4 RCS Reheat Duration Thresholds Cold SD/

Refueling 5 6

• If an RCS heat removal system is in operation within this time frame and RCS CA3.1 5 6 CU3.1 System temperature is being reduced the EAL is not applicable Malfunct. UNPLANNED increase in RCS temperature to > 200°F UNPLANNED increase in RCS temperature to > 200°F None Containment Heat-up for > Table C-4 duration (Notes 1, 10) (Note 10)

RCS RCS Status Closure Status Duration OR Temp. UNPLANNED RCS pressure increase > 10 psig (This EAL CU3.2 5 6 Intact (but not N/A 60 min.

• does not apply during water-solid plant conditions)

REDUCED INVENTORY) Loss of all RCS temperature and RCS level indication for 15 min. (Note 1)

Not intact established 20 min.

• OR REDUCED INVENTORY not established 0 min. Loss of required DC power for 15 minutes or longer CU4.1 5 6 None None Loss of < 107 VDC bus voltage indications on Technical Table C-5 Communications Methods Vital DC Specification required 125 VDC buses for 15 min. (Note 1)

Power System Onsite ORO NRC Loss of all onsite or offsite communications capabilities Gaitronics X Plant Radios X CU5.1 5 6 DEF Plant Emergency Dedicated Phones X Loss of all Table C-5 onsite communication methods None None Plant Telephone System X X X OR Loss of ENS (Red Phone) Line X X Loss of all Table C-5 ORO communication methods Comm.

Back-Up Radio System X OR Sentry Notification System X Loss of all Table C-5 NRC communication methods Hazardous event affecting a SAFETY SYSTEM needed for the Table C-6 Hazardous Events current operating mode Seismic event (earthquake) CA6.1 5 6 Internal or external FLOODING event The occurrence of any Table C-6 hazardous event Hazardous None High winds or tornado None strike AND EITHER: None Event Event damage has caused indications of degraded FIRE Affecting performance in at least one train of a SAFETY Safety EXPLOSION SYSTEM needed for the current operating mode Systems Other events with similar hazard The event has caused VISIBLE DAMAGE to a characteristics as determined by the SAFETY SYSTEM component or structure needed Emergency Coordinator for the current operating mode Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Note 10: Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal EIP-ZZ-00101, Addendum 1, Rev.[xx]

EAL Classification Matrix Page 3 of 3 Power Operation Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled Prepared for Ameren by: Operations Support Services, Inc. - 6/11/15