ML15197A251: Difference between revisions

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{{Adams
#REDIRECT [[ULNRC-06230, 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]]
| number = ML15197A251
| issue date = 07/16/2015
| title = 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
| author name = Maglio S A
| author affiliation = Ameren Missouri
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NMSS, NRC/NRR
| docket = 05000483, 07201045
| license number = NPF-030
| contact person =
| case reference number = NEI 99-01, Rev. 6, TAC MF4945, ULNRC-06230
| document type = Letter, Response to Request for Additional Information (RAI)
| page count = 278
| project = TAC:MF4945
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:W Ameren MISSOURI Callaway Plant *********************************************************************************************************************************************************************************************************
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July 16 , 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 ULNRC-06230 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. RENEWED FACILITY OPERATING LICENSE NPF-30 SUPPLEMENT TO 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)
 
==References:==
: 1) ULNRC-06143, "License Amendment Request for Emergency Action Level (EAL) Upgrade Adopting NRC-Endorsed NEI 99-01, Revision 6," dated October 2, 2014 (ADAMS Accession Number ML14275A435)
: 2) ULNRC-06227, "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)," dated July 6; 2015 In Reference 2, Ameren Missouri submitted a response to NRC staff requests for additional information (RAis) that were needed to complete the staffs review of a license amendment request (submitted as Reference
: 1) to upgrade the Emergency Action Level scheme associated with the Radiological Emergency Response Plan (RERP) for Callaway Unit 1 by adopting Nuclear Energy Institute (NEI) 99-01 , Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors." ******************************************************************************************************
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PO Box 620 Fulton, MD 65251 AmerenMissouri.com Alliance .............. . ............ . . ............ . ************** . ............ . ..............
ULNRC-06230 July 16, 2015 Page2 As a supplement to the Reference 2 RAI response, corrections and additions to the previously submitted material are provided in the Attachments to this letter. Attachment 1 provides a corrected RAI response matrix, wherein it is noted for the response to RAI 30 that EAL HS5.1 is applicable in Modes 1, 2, 3, 4, 5 and 6. Attachment 2 contains an amended summary ofEAL changes that were not associated with RAI responses, including changes to the Bases for EALs CALl, CGl.l, SUI. I, SAl. I and SS 1.1 that had been omitted from the Reference 2 submittal.
Attachment 3 provides a copy of the proposed RERP Technical Bases document that has been marked-up to reflect the changes made to the original submitted in Reference
: 1. Also, for information, Attachment 4 provides a copy of the proposed EAL wall charts that shows the corrected revision date (i.e., 7 /2115) in the chart footers. The Callaway Onsite Review Committee has approved the proposed changes to the RERP and its Technical Bases. In addition, in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," Section (b )(1), a copy of this letter is being provided to the designated Missouri State official.
This letter 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 16, 2015 A.. 0 Scott A. Maglio Manager, Regulatory Affairs JPK/nls Attachments:
: 1) Corrected Response to Request for Additional Information (RAI) Emergency Action Level (EAL) Scheme Change 2) Corrected Summary of EAL Changes NOT Associated with RAI Responses
: 3) Revised Callaway NEI 99-01 Revision 6 EAL Technical Bases (Marked-up Copy) 4) Corrected Callaway NEI 99-01 Revision 6 EAL Wall Charts (Information Only)
ULNRC-06230 July 16, 2015 Page 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 7 6011-4511 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 of Nuclear 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-06230 July 16, 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-06230 Corrected Response to Request for Additional Information (RAI) Emergency Action Level (EAL) Scheme Change 20 Pages RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 1 of 20 RAI-CP SECTION/EAL Question CP Response 01 4.3 Section 4.3, "Instrumentation Used for EALs," to NEI 99-01, Revision 6, states "Scheme developers should ensure that specific values used as EAL setpoints are within the calibrated range of the referenced instrumentation."  Please confirm that 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. CP has confirmed that 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. 02 1.0 In regards to Section 1, "Purpose," of the proposed EAL Technical Basis:
: a. Section 4.6, "Basis Document," to NEI 99-01, 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."
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.
: a. Deleted: "It should be used to facilitate review of the Callaway EALs and provide historical documentation for future reference." 
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 2 of 20 RAI-CP SECTION/EAL Question CP Response
: b. Section 4.6, "Basis Document," to NEI 99-01, Revision 6, states "Because the information in a basis document can affect emergency classification decision-making-" Therefore, the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). Please incorporate information 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 AOP [Abnormal Operating Procedure] and EOP [Emergency Operating Procedure]
Setpoints/Criteria," to NEI 99-01, Revision 6, states "As reflected in the generic guidance, the criteria/values used in several EALs and fission product barrier thresholds may be drawn from a plant's 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.
: b. The following has been added to Section 1.0 Introduction of the Technical Bases document:
"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)."
: c. The following has been added to Section 1.0 Introduction of the Technical Bases document:
"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).
"
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 3 of 20 RAI-CP SECTION/EAL Question CP Response 03 2.1 Sections 2.1, "Background," and 4.0, "References," of the proposed EAL Technical Basis reference an incorrect ADAMS Accession Number (ML110240324). Please verify that the proposed EAL Technical Basis is consistent with NRC endorsed guidance and appropriate ADAMS Accession number is referenced. Revised the referenced ADAMS Accession No. to ML12326A805. 04 2.1 For Sections 2.1, "Background," and 4.0, "References," of the proposed EAL Technical Basis, please provide ADAMS Accession Number that references the endorsed version of NEI 99-01, Revision 6 (ML12326A805). Revised the referenced ADAMS Accession No. to ML12326A805. 05 2.5 Section 2.5, "Technical Basis Information," of the proposed EAL Technical Basis includes a Plant-Specific basis section, in addition to a Generic basis section. Considering that the EAL Technical Basis is provided to support proper emergency classification decision making, please explain why a Generic basis section is provided or revise accordingly. Separate site-specific and generic bases were identified within the EAL bases to facilitate NRC review. These bases sections have now been combined into a single bases section for each EAL.
(Throughout document)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 4 of 20 RAI-CP SECTION/EAL Question CP Response 06 2.6 Section 2.6, "Operating Mode Applicability," of the proposed EAL Technical Basis contains a brief discussion concerning EAL classification during mode changes. However, this discussion is not as clear as that provided in NRC endorsed guidance.
Please justify the omission of significant portions of Section 5.4, "Consideration of Mode Changes During Classification," of NEI 99-01, Revision 6, or revise accordingly. Revised the cited section to read consistent with Section 5.4 of the generic guidance:
"The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher."
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 5 of 20 RAI-CP SECTION/EAL Question CP Response 07 3.1.1 3.1.2 In regards to Section 3 of the proposed EAL Technical Basis:
Section 3.1.1, "Classification Timeliness," includes a reference to NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants," but does not include a discussion, as provided by NEI 99-01, Revision 6, Section 5.2, Classification Methodology," addressing "[w]hen 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'."  Please justify excluding this information or revise accordingly.
Section 3.1.2, "Valid Indications," does not include statement, "[t]he validation of indications should be completed in a manner that supports timely emergency declaration," as provided by NEI 99-01, Revision 6, Section 5.1, "General Considerations."  Please justify excluding this information or revise accordingly.
 
Added the cited wording to Section 3.1.1: "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."
 
Added the cited statement to Section 3.1.2: "The validation of indications should be completed in a manner that supports timely emergency declaration,"
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 6 of 20 RAI-CP SECTION/EAL Question CP Response 08 5.0 Appendix B, "Definitions," to NEI 99-01, Revision 6, provides definitions for key terms necessary for overall understanding of the NEI 99-01 emergency classification scheme. For Section 5.1, "Definitions," please revise accordingly to add definitions for the following or justify excluding:
* CONFINEMENT BOUNDARY,
* 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 SPENT FUEL STORAGE INSTALLATION to eliminate redundancy. Added the following definitions to Section 5.1:
* Confinement Boundary,
* Emergency Action Level
* Emergency Classification Level
* Fission Product Barrier Threshold
* Initiating Condition.
 
Deleted the duplicate ISFSI definition from Section 5.1. 09 6.0 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 included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference matrix. b. Callaway emergency action level CG1.1 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 included in the Callaway to NEI 99-01, 
: a. RA2.3 added to EAL Cross-Reference.
: b. Revised EAL Cross-Reference to cite correct reference. 
: c. CG1.2 added to EAL Cross-Reference.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 7 of 20 Revision 6, EAL Cross-Reference. d. Callaway emergency action level SU 4.1 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. The Callaway to NEI 99-01, Revision 6, EAL Cross-Reference includes Callaway EAL SU4.2. EAL SU4.2 could not be located in the Callaway EAL basis document. f. Callaway emergency action level SU8.1 is not included in the Callaway to NEI 99-01, Revision 6, EAL Cross-Reference matrix. g. The Callaway to NEI 99-01, Revision 6, EAL 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 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 Callaway to NEI 99-01, Revision 6, EAL Cross-Reference.  (Note:  The "E" designation is correct.) 
: d. Corrected cited reference.
: e. Deleted SU4.2 from EAL Cross-Reference. Generic SU 3 example 1 is not implemented at Callaway.
: f. Added SU8.1 to EAL Cross-Reference
: g. Corrected to read "SA9.1"
: h. See "g" above.
: i. Corrected designation to read "E"
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 8 of 20 RAI-CP SECTION/EAL Question CP Response 10 RU1.1 For EAL RU1.1, it is not clear how a determination can be made that a "2 X Hi - Hi alarm" condition exists. Please provide justification that a value of two times the alarms identified in Table R-1, "Effluent Monitor Classification Thresholds," can be accurately determined in a timely and accurate manner. GT-RE-21B, GH-RE-10B & HB-RE18 alarms, Hi-Hi alarm setpoints, and monitor indications are displayed on the RM-11 in the Control Room. It is a simple matter of multiplying by the Hi-Hi alarm setpoint by 2 to get the EAL indicator value. If the RM-11 Hi-Hi alarm were to come in, the Control room Operator would monitor the parameter for EAL applicability. 11 RA1.2, RS1.2, RG1.2 For EALs RA1.2, RS1.2, and RG1.2, please explain why the proposed Note 3, which relates to "effluent flow past an effluent monitor," should be included for EALs that are based on dose assessments or revise accordingly. Deleted Note 3 applicability to RA1.2, RS1.2 and RG1.2. 12 RA1.1, RS1.1, RG1.1 For EALS RA1.1, RS1.1, and RG1.1, there was a substantial change from the previous to the proposed Table R-1 values. The provided calculations did not contain information that could be used to justify this change. Please provide justification that supports the changes in the Table R-1 values from the previous values to the current values. The NEI 99-01 Revision 5 Table R-1 calculation is documented in EPCI-08-01 (refer to Attachment 5). The software program MAGNEM was used. Selection of source terms for MAGNEM is limited compared to the current 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. 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-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 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 9 of 20 RAI-CP SECTION/EAL Question CP Response 14 Generic AA3/HA5 For NEI 99-01, Revision 6, EALs AA3 and HA5, CP is proposing two deviations. NEI Initiating Condition (IC) AA3 example 2, and HA5 example 1 will not be included because a review of CP normal operating and shutdown procedures by Operations Subject Matter Experts concluded that there are no areas external to the Main Control Room that require access to perform a normal plant shutdown and cooldown to Cold Shutdown conditions. a. Please verify that all required manipulations to shut down the plant and enter shutdown cooling can be performed from the Main 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.
A review of all procedures associated with down power from 100% to Mode 5 (cold shutdown) reviewing field actions that have to be taken. After review, there are no rooms which need to be accessed to shut the plant down to Mode 5. RHR shutdown cooling does not have to be placed in service because CP can cool down to Mode 5 using ASD, steam dumps and MSIV bypasses. Closing the breakers for SI Accumulators and RHR loop suctions valves would not be required.
The end result was that a shutdown to cold shutdown can be accomplished from the Control Room alone, without additional actions being performed. The Control Room Ventilation System provides adequate protection from external hazardous gases.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 10 of 20 RAI-CP SECTION/EAL Question CP Response 15 RU2.1 RA2.1 RA2.2 RA2.3 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." a. Please provide site-specific level indications for EAL RU2.1 that could be used to support timely and accurate assessments; include applicable mode availability for this level instrumentation. b. Please justify excluding the NEI 99-01, Revision 6, EAL AA2 guidance that relates to RA2.1, RA2.2, and RA2.3 applicability or revise accordingly. c. Please verify that RA2.1 should be an Alert and revise accordingly.
: a. Added "(EC LI-0039A, EC LI-0039B, local observation of SFP level)" as site-specific SFP level indication to RU2.1. 
: b. Added the cited applicability statement to the RA2.1 and RA2.2 bases: "This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1.1."  c. Corrected typo to read "Alert."
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 11 of 20 RAI-CP SECTION/EAL Question CP Response 16 RA2.2 For EAL RA2.2, the logic was changed from NEI 99-01, Revision 6, guidance that uses an increase in radiation monitor readings to determine that irradiated fuel has been damaged to a proposed logic that requires the operator to know that damage has occurred to irradiated fuel AND an there is an increase in radiation monitor indications. Please develop EAL RA2.2 per NEI 99-01, Revision 6, as endorsed or provide further justification for this deviation. RA2.2 has been revised to read:
"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 mR/hr
* Fuel Pool Bridge Crane OR Spent Fuel Pool Area radiation monitor (SD-RE-37 or 38) > 30 mR/hr" 17 EU1.1 For EAL EU1.1, please explain why symbols were used rather than spelling out "gamma" and "neutron" or revise accordingly. Revised EU1.1 by replacing the gamma and neutron symbols with the terms "gamma" and "neutron."
Reworded the EAL as follows:
Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading > EITHER of the following:
* 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.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 12 of 20 RAI-CP SECTION/EAL Question CP Response 18 CA1.1 For EAL CA1.1, a BBLI-53 A/B level of 0 inches is provided as an indication that RCS level is lower than the bottom of the RCS hot leg. The Callaway Basis provides that BBLI-53A/B cannot sense level changes in the Reactor Vessel below the 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. BBLI-53 A/B indications are trended and displayed when providing RCS level indication. Comparison between BBLI-53A and BBLI-53B, along with their trending together, provides adequate indication of failure of a channel. (No document change) 19 N/A Omitted N/A 20 CS1.1 CS1.2 For EAL CS1.1 and CS1.2, the logic was changed from NEI 99-01, Revision 6, guidance without justification. As changed, the EAL appears vague and interpretive. Please develop EAL CS1.1 and CS1.2 per NEI 99-01, Revision 6, as endorsed, or provide further justification for deviation. Both CS1.1 and CS1.2 wording is consistent with the generic guidance intent. The status of containment closure modifies the level threshold consistent with the generic guidance. 
(No document change) 21 CS1.3 CG1.2 For EALs CS1.3 and CG1.2, CP did not include "of sufficient magnitude to indicate core uncovery" to the unplanned increase in any sump/tank level to the IC wording. Additionally, CP added "Visual observation of UNISOLABLE RCS leakage" to the IC wording. As proposed, EALs CS1.3 and 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. Added "of sufficient magnitude to indicate core uncovery" to sump/tank level increases in CS1.3 and CG1.2.
 
Deleted visual observation of unisolable RCS leakage from CS1.3 and CG1.2.
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 13 of 20 RAI-CP SECTION/EAL Question CP Response 22 CG1.2 For EAL CG1.2 and the Containment Fission 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.
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. Deleted reference to CA-3 in CG1.2.
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 14 of 20 RAI-CP SECTION/EAL Question CP Response 23 CU2.1, CA2.1, SA1.1, SS1.1, SG1.1, SG1.2 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
: b. Please justify the addition of "Additional sources of offsite power are available from diesel generators such as the Alternate Emergency Power Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes" to the Callaway Basis as this statement could potentially be applied to power supplies not listed on Table C-3 or revise according to NRC endorsed guidance.
: a. N/A
: b. Per the developer notes: "The EAL and/or Basis section may specify use of a non-safety-related power source provided that operation of this source is controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines). Such power sources should generally meet the "Alternate ac source" definition provided in 10 CFR 50.2" The Alternate Emergency Power Supply (AEPS) was built for, and has been demonstrated to be capable of carrying one of the two Safety Related 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)
 
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 15 of 20 RAI-CP SECTION/EAL Question CP Response 24 CU3.1 CA3.1 For EAL CU3.1 and CA3.1, please explain how 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 result in potential misclassification for an event other than a loss of decay heat removal that leads to an unplanned RCS temperature and/or RCS pressure rise. Please provide justification or revise accordingly consistent with endorsed guidance Deleted "-due to the loss of decay heat removal capability-" from CU3.1.
 
Deleted "-due to a loss of RCS cooling-" from CA3.1. Added the following omitted generic EAL wording: "(This EAL does not apply during water-solid plant conditions.)"  25 CU5.1 SU7.1 For EALs CU5.1 and SU7.1, the Sentry Notification System is provided as an offsite response organization (ORO) communication method for the electronic transmission of a notification form to the OROs. Please provide reference to specific section of the site emergency plan that that identifies the Sentry Notification System as a means of timely notification to OROs for a spectrum of potential event responses or revise accordingly. Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2.4. "The Sentry System provides a means of performing required ORO emergency notifications in a timely manner (within 15 minutes of classification).  " When each ORO acknowledges receipt of a Sentry notification, Sentry provides Callaway Plant with positive confirmation that ORO received the message.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 16 of 20 RAI-CP SECTION/EAL Question CP Response 26 CA3.1 For EAL CA3.1, Note 10: "Begin monitoring hot condition EALs concurrently," was added to the provided EAL Technical Basis. It is not clear to the staff how Note 10 would be applied during an UNPLANNED increase in RCS temperature event.
Please provide justification for this difference or revise accordingly. Revised Note 10 to read:
"Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal
." This note re-enforces the implementation guidance in NEI 99-01 that states: "The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. 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) 27 CA6.1 SA9.1 For EAL CA6.1 and SA9.1, the Callaway Basis discussion for seismic events refers to a discussion under EAL HU2.1. Please include the discussion on seismic events in the EAL CA6.1 and SA9.1 Callaway Basis or provide justification for not including the discussion as this could impact the timeliness of event assessment. Added the applicable portion of the seismic event discussion in HU2.1 to the CA6.1 and SA9.1 bases.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 17 of 20 RAI-CP SECTION/EAL Question CP Response 28 HU2.1 For EAL HU2.1, the proposed EAL may not be consistent with from NEI 99-01, Revision 6, guidance, which provides that "site-specific indication that a seismic event met or exceeded OBE [operating basis earthquake] limits" should be based on the indications, alarms, and displays of site-specific monitoring equipment. The 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. The HU2.1 wording was intended for the Unusual Event classification to be driven by receipt of seismic activity annunciator 98D (OBE) in the control room and not by actions performed in the seismic event AOP (OTO-SG-00001). Revised HU2.1 to read" "Seismic event > OBE as indicated by Seismic Activity, Annunciator 98D." 29 HU3.2 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. a. Please explain how the statement in EAL HU3.2 that limits flooding areas of concern will 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 references an EAL that is applicable in lower modes. 
: a. Deleted bases statement related to internal flooding areas. 
: b. Added reference to SA9.1 in addition to CA6.1 in the HU3.2 bases.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 18 of 20 RAI-CP SECTION/EAL Question CP Response 30 HS5.1 For EAL HS5.1, please consider the following or provide an explanation how this EAL can be consistently applied:
* Addition of operating mode specificity to the listed safety functions to preclude event classification when these safety functions are no longer needed in accordance with site technical specifications; and
* Including a "Clock" start time in the Callaway Basis discussion.
 
1st bullet - Revised to restrict mode applicability of HS5.1 to Modes 1, 2, 3, 4, 5, and 6. None of the three listed safety functions are required when the reactor vessel is defueled.
Restricted reactivity control safety function to Modes 1, 2 and 3 only since, by definition, the reactor has adequate shutdown margin while in Modes 4 and 5.
 
2nd bullet - Added the following to the bases regarding "clock" start time: "For the purpose of this EAL the 15 minute clock starts when the last licensed operator leaves the Control Room."
31 SU3 [SU4.1] For NEI 99-01, Revision 6, EAL SU3.1, CP does not provide an EAL that uses site-specific radiation monitor(s). Please provide additional justification that a Callaway EAL cannot be developed consistent with endorsed guidance or revise accordingly. Callaway does not have a site-specific radiation monitor correlation that supports identifying a monitor reading that correspond to TS coolant activity limits.
The associated EAL is based on whether or not Technical Specification limits for RCS activity have been exceeded. Callaway's Technical Specifications include limits for Iodines governed by Dose Equivalent Iodine (DEI) and noble gases governed by Dose Equivalent Xenon (DEXe). DEI and DEXe both represent weighted sums of measured isotopic concentrations. Determination 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)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 19 of 20 RAI-CP SECTION/EAL Question CP Response 32 SU4.1 For EAL SU4.1, please explain why the proposed wording is different from the NEI 99-01, Revision 6, guidance which clearly states "sample analysis indicates that-," or revise accordingly. Revised SU4.1 to read: "Sample analysis indicates RCS activity > Technical Specification Section 3.4.16 limits" 33 SU5.1 For EAL SU5.1, please explain how timely declaration can be performed without reliance on a potentially time consuming "manual" method of performing an RCS inventory balance or revise the Callaway Basis accordingly. Callaway does not rely on a manual method of performing an RCS inventory balance. As stated in the bases: "Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage. The Personal 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, 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) 34 SU6.1 For EAL SU6.1, please provide a justification for including a "subsequent automatic trip" to the EAL condition or revise accordingly The words "subsequent automatic trip" were added to both SU6.1 and SU6.2 to address the condition where an automatic trip signal other than the initial 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 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)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) EMERGENCY ACTION LEVEL (EAL) SCHEME CHANGE Callaway Plant, Unit 1 (CP) DOCKET NO. 50-483 Page 20 of 20 RAI-CP SECTION/EAL Question CP Response 35 SU6.1, SU6.2, SA6.1, SS6.1 For EALs SU6.1, SU6.2, SA6.1, and SS6.1, please provide further justification as to why greater than or equal to five percent reactor power was added or revise accordingly.  (Note:
Westinghouse EOPs do not solely rely on Reactor Power level to determine the status of reactor criticality.) The method used to determine that the reactor is shutdown following a reactor trip, for the purposes of emergency classification, is consistent with the Callaway EOPs (E-0), i.e., indication of reactor power < 5%. This is also the power level that defines power operation in the Technical Specifications. As specified in the generic developers guidance:  "Developers may include site-specific EOP criteria indicative of a successful 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) 
 
Attachment 2 to ULNRC-06230 Corrected Summary of EAL Changes NOT Associated with RAI Responses 1 Page 
 
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 Abbreviations/Acronyms ODCM - Removed the "-" from Off-site for consistency throughout the document. 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". CA1.1 In the Basis section, corrected "6.5%" to "the specified level." CG1.1 In the Basis and Callaway Basis Reference(s) sections, deleted references to CA-3. SU1.1,SA1.1 and SS1.1  In the Basis section, deleted the sentence "The 15-minute interval was selected as a threshold to exclude transient or momentary power losses." 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 B.1. Potential Loss In the Basis, deleted duplicate "indicates" in first sentence. RCS C Potential Loss In the Category line, corrected "
B. CMT Radiation/RCS Activity" to "
C. CMT Radiation/RCS Activity". RCS E.1 Potential Loss In the Basis section, corrected "Emergency Director" to "Emergency Coordinator". 
 
Attachment 3 to ULNRC-06230 Revised Callaway NEI 99-01 Revision 6 EAL Technical Bases (Marked-up Copy) 246 Pages
 
Page 1 of 245 INFORMATION USE
 
EIP-ZZ-00101 ADDENDUM 2 Revision xxx Draft 07/02/15 (GLJ)
 
Page 2 of 245 INFORMATION USE TABLE OF CONTENTS SECTION PAGE 1.0PURPOSE ................................................................................................................................. 32.0DISCUSSION ............................................................................................................................ 32.1Background ............................................................................................................................ 32.2Fission Product Barriers ......................................................................................................... 42.3Fission Product Barrier Classification Criteria ....................................................................... 42.4EAL Organization ................................................................................................................... 52.5Technical Bases Information .................................................................................................. 62.6Operating Mode Applicability ................................................................................................. 83.0GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ................................................ 93.1General Considerations ......................................................................................................... 93.2Classification Methodology .................................................................................................. 1
 
==04.0REFERENCES==
........................................................................................................................ 134.1Developmental ..................................................................................................................... 134.2Implementing ........................................................................................................................ 135.0DEINITIONS, ACRONYMS & ABBREVIATIONS ................................................................... 146.0Callaway TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ................................................. 207.0ATTACHMENTS ..................................................................................................................... 241 Emergency Action Level Technical Bases ................................................................. 24 Category R Abnormal Rad Release / Rad Effluent ........................................... 25 Category E ISFSI .............................................................................................. 64 Category C Cold Shutdown / Refueling System Malfunction ........................... 67 Category H Hazards ........................................................................................ 112 Category S System Malfunction ..................................................................... 147 Category F Fission Product Barrier Degradation ........................................... 190 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ...................................................................................................... 195
 
Page 3 of 245 INFORMATION USE
 
===1.0 PURPOSE===
 
This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Callaway Energy Center (Callaway
) Decision-makers responsible for implementation of EIP-ZZ-00101 Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to 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 Background===
 
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).
In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
* Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).
* Simplifying the fission product barrier EAL threshold for a Site Area Emergency. Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref. 4.1.1), Callaway conducted an EAL implementation upgrade project that produced the EALs discussed herein. ). It should be used to facilitate review of the Callaway EALs and provide historical documentation for future reference. 110240324
 
Page 4 of 245 INFORMATION USE
 
===2.2 Fission===
Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.
The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System (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 includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency 2.3 Fission Product Barrier Classification Criteria  The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier 
 
Page 5 of 245 INFORMATION USE 2.4 EAL Organization The Callaway EAL scheme includes the following features:
* Division of the EAL set into three broad groups:
o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.
o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
* Within each group, assignment of EALs to categories and subcategories: Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The 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 subcategories are listed below.   
 
Page 6 of 245 INFORMATION USE EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode:
R - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting Plant Safety
 
E - ISFSI 1 - Security 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Control Room Evacuation 6 - Emergency Coordinator Judgment 1 - Confinement Boundary 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 Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:
C - Cold Shutdown / Refueling System Malfunction 1 - RCS Level 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.
 
===2.5 Technical===
Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, I and F) and EAL subcategory. A summary explanation 
 
Page 7 of 245 INFORMATION USE of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier: 1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, E or F) 2. Second character (letter): The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1). 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1). Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix 
 
Page 8 of 245 INFORMATION USE Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling, D - Defueled, or 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. 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 NEI 99-01 Rev. 6.
Callaway Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.8) 1 Power Operation Keff  0.99 and reactor thermal power > 5% 2 Startup Keff  0.99 and reactor thermal power  5% 3 Hot Standby Keff < 0.99 and average coolant temperature  350&#xba;F 4 Hot Shutdown Keff < 0.99 and average coolant temperature 350&#xba;F > Tavg > 200 &#xba;F and at least 53 of 54 reactor vessel head closure bolts fully tensioned 5 Cold Shutdown Keff < 0.99 and average coolant temperature  200&#xba;F 6 Refueling One or more reactor vessel head closure bolts are less than fully tensioned D Defueled All fuel 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 effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 
 
Page 9 of 245 INFORMATION USE The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs.
If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
 
Page 10 of 245 INFORMATION USE
 
===3.0 GUIDANCE===
ON MAKING EMERGENCY CLASSIFICATIONS
 
===3.1 General===
Considerations When making an emergency classification, the Emergency Coordinator must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds. 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.11). 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 valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's 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 should 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 (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 indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
 
====3.1.3 Imminent====
Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
 
====3.1.4 Planned====
vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and 
 
Page 11 of 245 INFORMATION USE 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 &sect; CFR 50.72 (ref. 4.1.4).
 
====3.1.5 Classification====
Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Coordinator with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Coordinator will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.
When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock."  For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.11). 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, whether at one unit or at two different units, a Site Area Emergency should be declared. There is no "additive" effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared. 
 
Page 12 of 245 INFORMATION USE Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2). 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Coordinator, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
 
====3.2.4 Emergency====
Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
 
====3.2.5 Classification====
of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.
 
====3.2.6 Classification====
of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some 
 
Page 13 of 245 INFORMATION USE transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.
In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR &sect; 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3). 
 
Page 14 of 245 INFORMATION USE
 
==4.0 REFERENCES==
 
===4.1 Developmental===
4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 &sect; CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 &sect; CFR 50.73 License Event Report System
 
====4.1.6 Drawing====
8600-X-88100 Property-Site Layout Owner Controlled Area and Surrounding Area 4.1.7 Callaway UFSAR 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-01 Interim Staff Guidance, 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 4.2 Implementing 4.2.1 EIP-ZZ-00101 Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to Callaway EAL Comparison Matrix 4.2.3 Callaway EAL Matrix 110240324
 
Page 15 of 245 INFORMATION USE 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 or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the Callaway ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC). Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
As applied to 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 Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Unusual Event (UE), Alert, Site Area Emergency (SAE) and General Emergency (GE). EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires 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 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. 
 
Page 16 of 245 INFORMATION USE 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.
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.   
 
Page 17 of 245 INFORMATION USE 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.
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 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 determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Impede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Independent Spent Fuel Storage Installation (ISFSI)
A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating Condition An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Maintain Take appropriate action to hold the value of an identified parameter within specified limits. Independent Spent Fuel Storage Installation (ISFSI)&#xb6; A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.&#xb6;
 
Page 18 of 245 INFORMATION USE Owner Controlled Area (OCA) The fenced area contiguous to the Protected Area, designated by AmerenUE (Callaway Plant) to be controlled for security purposes (ref 4.1.6).
Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
Protected Area (PA) An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. (ref. 4.1.7).
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). 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.
Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.   
 
Page 19 of 245 INFORMATION USE 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 functions needed for protection of the public or hostile actions that result 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 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 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 (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 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 material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Visible Damage Damage to a component or structure 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 component or structure. 
 
Page 20 of 245 INFORMATION USE 5.2 Abbreviations/Acronyms
&deg;F ......................................................................................................... Degrees Fahrenheit &deg; .............................................................................................................................. Degrees AC ......................................................................................................... Alternating Current ATWS ........................................................................ Anticipated Transient Without Scram Callaway ...................................................................................... Callaway Energy Center CDE ........................................................................................ Committed Dose Equivalent CFR ....................................................................................... Code of Federal Regulations CMT ................................................................................................................ Containment CSFST ........................................................................Critical Safety Function Status Tree DBA ..................................................................................................Design Basis Accident DC ................................................................................................................. Direct Current EAL .............................................................................................. Emergency Action Level ECCS ............................................................................. Emergency Core Cooling System ECL ................................................................................... Emergency Classification Level EOF .................................................................................... Emergency Operations Facility EOP ................................................................................ Emergency Operating Procedure EPA ................................................................................ Environmental Protection Agency ERG ................................................................................. Emergency Response Guideline EPIP ................................................................. Emergency Plan Implementing Procedure ESF .......................................................................................... Engineered Safety Feature ESW .......................................................................................... Emergency 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 ................. Individual Plant Examination of External Events (Generic Letter 88-20)
Keff .......................................................................... 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, mRem, mrem, mREM  ............................................... milli-Roentgen Equivalent Man MSL .......................................................................................................... Main Steam Line MW ....................................................................................................................... Megawatt NEI ................................................................................................ Nuclear Energy Institute NESP .................................................................... National Environmental Studies Project 
 
Page 21 of 245 INFORMATION USE 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 Square Inch Gauge R ........................................................................................................................... Roentgen RCC ............................................................................................. Reactor Control Console RCS ............................................................................................. Reactor Coolant System Rem, rem, REM  ........................................................................ Roentgen Equivalent Man RETS .......................................................... Radiological Effluent Technical Specifications RPS .......................................................................................... Reactor Protection System R(P)V ........................................................................................ Reactor (Pressure) Vessel RVLIS ................................................................... Reactor Vessel Level Indicating System SAR ................................................................................................. Safety Analysis Report SBO .......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SG ............................................................................................................ Steam Generator SI ................................................................................................................. Safety Injection ODCM .............................................................................. Selected Licensee Commitment SPDS ............................................................................ Safety Parameter Display System SRO ............................................................................................. Senior Reactor Operator SSF ................................................................................................. Safe Shutdown Facility TEDE ................................................................................ Total Effective Dose Equivalent TOAF ..................................................................................................... Top of Active Fuel TSC ............................................................................................ Technical Support Center WOG .................................................................................... Westinghouse Owners Group
-
 
Page 22 of 245 INFORMATION USE 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 NEI 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 99-01 Rev. 6 EAL IC Example EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 
 
Page 23 of 245 INFORMATION USE Callaway NEI 99-01 Rev. 6 EAL IC Example EAL CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 2
 
Page 24 of 245 INFORMATION USE Callaway NEI 99-01 Rev. 6 EAL IC Example EAL HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU6.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA6 1 HA6.1 HA7 1 HS1.1 HS1 1 HS5.1 HS6 1 HS6.1 HS7 1 HG1.1 HG1 1 HG6.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 1SU4.2SA8
 
Page 25 of 245 INFORMATION USE Callaway NEI 99-01 Rev. 6 EAL IC Example EAL SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 IU1
 
Page 26 of 245 INFORMATION USE
 
===7.0 ATTACHMENTS===
 
===7.1 Attachment===
1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Loss/Potential Loss Matrix and Basis
 
ATTACHMENT 1 EAL Bases  Page 27 of 245 INFORMATION USE Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Events of this category pertain to the following subcategories: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
: 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
: 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
ATTACHMENT 1 EAL Bases  Page 28 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for  60 min.  (Notes 1, 2, 3)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Gaseous Unit Vent GT-RE-21B 6.59E+7 &#xb5;Ci/sec 6.59E+6 &#xb5;Ci/sec 6.59E+5 &#xb5;Ci/sec 2 X Hi-Hi alarm ASD Monitors (A/B/C/D) AB-RE-111/112/113/114 12 mR/hr 1.2 mR/hr ---- ---- TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ---- Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste Discharge HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm
 
Mode Applicability:
All ATTACHMENT 1 EAL Bases  Page 29 of 245 INFORMATION USE Definition(s):
None Basis: The column "UE" 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). The effluent monitor Hi-Hi alarm setpoints correspond to the Hi alarm (red) setpoint as displayed on RM-11.
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
Escalation of the emergency classification level would be via IC RA1. Callaway Basis Reference(s):
: 1. APA-ZZ-01003 Callaway Plant Offsite Dose Calculation Manual Section 2.2.3 2. FSAR Section 16.11.1.3, Radioactive Effluent Monitoring Instrumentation LCO 3. EPCI 11402 EAL Table R-1 Calculations 
: 4. NEI 99-01 AU1
 
Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 30 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
EAL:
RU1.2  Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 
> 2 x ODCM limits for  60 min. (Notes 1, 2)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:
All Definition(s):
None Basis: None This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RA1. Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 31 of 245 INFORMATION USE Callaway Basis Reference(s):
: 1. APA-ZZ-01003 Callaway Plant Offsite Dose Calculation Manual Section 2.2.3 2. NEI 99-01 AU1 ATTACHMENT 1 EAL Bases  Page 32 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.1 Alert Reading on any Table R-1 effluent radiation monitor > column "ALERT" for  15 min. (Notes 1, 2, 3, 4)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Gaseous Unit Vent GT-RE-21B 6.59E+7 &#xb5;Ci/sec 6.59E+6 &#xb5;Ci/sec 6.59E+5 &#xb5;Ci/sec 2 X Hi-Hi alarm ASD Monitors (A/B/C/D) AB-RE-111/112/113/114 12 mR/hr 1.2 mR/hr ---- ---- TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ---- Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste Discharge HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm
 
Mode Applicability:
All ATTACHMENT 1 EAL Bases  Page 33 of 245 INFORMATION USE Definition(s):
None Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 10 mRem TEDE
* 50 mRem CDE Thyroid The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 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 un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS1. Callaway Basis Reference(s):
: 1. EPCI 11402 EAL Table R-1 Calculations
: 2. NEI 99-01 AA1
 
Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 34 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: Dose assessments are performed by computer-based method (ref. 1, 2)
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS1. s 3,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.&#xb6;Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 35 of 245 INFORMATION USE Callaway Basis Reference(s):
: 1. EIP-ZZ-01211 Accident Dose Assessment
: 2. EPCI 11402 EAL Table R-1 Calculations
: 3. NEI 99-01 AA1
 
ATTACHMENT 1 EAL Bases  Page 36 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE  EAL:
RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (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 un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 37 of 245 INFORMATION USE stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS1. Callaway Basis Reference(s):
: 1. APA-ZZ-01003 Callaway Plant Offsite Dose Calculation Manual Section 2.2.3 2. NEI 99-01 AA1
 
ATTACHMENT 1 EAL Bases  Page 38 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE  EAL:
RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:  Closed window dose rates > 10 mR/hr expected to continue for  60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (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 un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 39 of 245 INFORMATION USE Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS1. Callaway Basis Reference(s):
: 1. EIP-ZZ-00211, Field Monitoring
: 2. NEI 99-01 AA1
 
ATTACHMENT 1 EAL Bases  Page 40 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.1  Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for  15 min.  (Notes 1, 2, 3, 4)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Gaseous Unit Vent GT-RE-21B 6.59E+7 &#xb5;Ci/sec 6.59E+6 &#xb5;Ci/sec 6.59E+5 &#xb5;Ci/sec 2 X Hi-Hi alarm ASD Monitors (A/B/C/D) AB-RE-111/112/113/114 12 mR/hr 1.2 mR/hr ---- ---- TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ---- Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste Discharge HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm
 
Mode Applicability:
All Definition(s):
None ATTACHMENT 1 EAL Bases  Page 41 of 245 INFORMATION USE Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
* 100 mRem TEDE
* 500 mRem CDE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1). This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RG1. Callaway Basis Reference(s):
: 1. EPCI 11402 EAL Table R-1 Calculations
: 2. NEI 99-01 AS1 CallawayNEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 42 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or  500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: Dose assessments are performed by computer-based method (ref. 1, 2)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. 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. Callaway Basis Reference(s):
s 3,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.&#xb6;Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 43 of 245 INFORMATION USE
: 1. EIP-ZZ-01211 Accident Dose Assessment
: 2. EPCI 11402 EAL Table R-1 Calculations
: 3. NEI 99-01 AS1
 
ATTACHMENT 1 EAL Bases  Page 44 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:  Closed window dose rates > 100 mR/hr expected to continue for  60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC RG1. Callaway NEI 99-01Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 45 of 245 INFORMATION USE Callaway Basis Reference(s):
: 1. EIP-ZZ-00211, Field Monitoring
: 2. NEI 99-01 AS1
 
ATTACHMENT 1 EAL Bases  Page 46 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for  15 min.  (Notes 1, 2, 3, 4) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Gaseous Unit Vent GT-RE-21B 6.59E+7 &#xb5;Ci/sec 6.59E+6 &#xb5;Ci/sec 6.59E+5 &#xb5;Ci/sec 2 X Hi-Hi alarm ASD Monitors (A/B/C/D) AB-RE-111/112/113/114 12 mR/hr 1.2 mR/hr ---- ---- TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ---- Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste Discharge HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm
 
Mode Applicability:
All Definition(s):
None Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either: Callaway ATTACHMENT 1 EAL Bases  Page 47 of 245 INFORMATION USE
* 1000 mRem TEDE
* 5000 mRem CDE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1). This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Callaway Basis Reference(s):
: 1. EPCI 11402 EAL Table R-1 Calculations
: 2. NEI 99-01 AG1
 
NEI 99-01Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 48 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or  5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: Dose assessments are performed by computer-based method (ref. 1, 2)
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Callaway Basis Reference(s):
: 1. EIP-ZZ-01211 Accident Dose Assessment s 3,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.&#xb6;Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 49 of 245 INFORMATION USE
: 2. EPCI 11402 EAL Table R-1 Calculations
: 3. NEI 99-01 AG1
 
ATTACHMENT 1 EAL Bases  Page 50 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:  Closed window dose rates > 1,000 mR/hr expected to continue for  60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation. (Notes 1, 2)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:
All Definition(s):
SITE BOUNDARY - 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 10CFR100.
Basis: EIP-ZZ-00211, Field Monitoring s provides guidance for emergency or post-accident radiological environmental monitoring (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 Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Callaway Basis Reference(s):
Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 51 of 245 INFORMATION USE
: 1. EIP-ZZ-00211, Field Monitoring
: 2. NEI 99-01 AG1
 
ATTACHMENT 1 EAL Bases  Page 52 of 245 INFORMATION USE 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 (EC LI-0039A
, EC LI-0039B, local observation of SFP 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:
* SD-RE-34, Cask Handle Area Radiation
* SD-RE-35, New Fuel Storage Area Radiation
* 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
* SD-RE-41, Manipulator Crane Radiation Monitor
* SD-RE-42, Containment Building Radiation
* GT-RE-59 Containment High Area Radiation Monitor
* 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 PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.
 
ATTACHMENT 1 EAL Bases  Page 53 of 245 INFORMATION USE Basis: The low water level alarm in this EAL refers to the Spent Fuel Pool (SFP) low level alarm (Window Number 76D, SFP 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 SFP, 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 Technical Specification Section 3.7.15 (ref. 2) requires at least 23 ft of water above the Spent Fuel Pool storage racks. Technical Specification Section 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 pool, 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 inventory (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 cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. Callaway Basis Reference(s):
: 1. OTO-EC-00001 Loss of Spent Fuel Pool/Refuel Pool Level 2. Technical Specification Section 3.7.15 Fuel Storage Pool Water Level 3. Technical Specification Section 3.9.7 Refueling Pool Water Level 4. NEI 99-01 AU2 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 54 of 245 INFORMATION USE
 
ATTACHMENT 1 EAL Bases  Page 55 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All Definition(s):
REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.
Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL 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 EU1.1. This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RS1. Callaway Basis Reference(s):
: 1. OTO-EC-00001 Loss of Spent Fuel Pool/Refuel Pool Level 2. NEI 99-01 AA2 Unusual EventCallaway None.&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 56 of 245 INFORMATION USE
 
ATTACHMENT 1 EAL Bases  Page 57 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity 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 mR/hr
* 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 fuel (ref. 1, 2).
The bases for the SFP 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. Should 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 irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1.1. Escalation of the emergency would be based on either Recognition Category R or C ICs. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a Damage to irradiated fuel resulting in a release of radioactivity &#xb6;AND&#xb6;Any  radiation monitor indicationsCallaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 58 of 245 INFORMATION USE potential fuel damaging event (e.g., a fuel handling accident). Escalation of the emergency classification level would be via IC RS1. Callaway Basis Reference(s):
: 1. OTO-EC-00001 Loss of Spent Fuel Pool/Refuel Pool Level 2. OTO-KE-00001 Fuel Handling Accident
: 3. Calc. EPCI 98-01 Emergency Action Level Bases
: 4. Calc. HPCI 05-02 Gaseous and Liquid Radiation Monitor Setpoints 5. NEI 99-01 AA2
 
ATTACHMENT 1 EAL Bases  Page 59 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3 Alert Lowering of spent fuel pool level to 2031 ft. 1.25 in. (Level 2)
Mode Applicability:
All Definition(s):
None Basis: Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). 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 on EC-LI-0059B in the Control Room or EC-LI-0059A in the Auxiliary Building MG Set Room. Backup indication is also available on EC-LI-0060A in the Auxiliary Building MG Set Room.
This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.
Escalation of the emergency would be based on either Recognition Category R or C ICs. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC RS1. 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
 
Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 60 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:
RS2.1 Site Area Emergency Lowering of spent fuel pool level to 2022 ft. 1.25 in. (Level 3)
Mode Applicability:
All Definition(s):
None Basis: Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). For Callaway Plant SFP Level 3 has been set at a plant elevation 2022 ft. 1.25 in. (~11 in. above the top of the spent fuel racks) as indicated on EC-LI-0059B in the Control Room or EC-LI-0059A in the Auxiliary Building MG Set Room. Backup indication is also available on EC-LI-0060A in the Auxiliary Building MG Set Room.
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC AG1 or RG2. 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 AS2 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 61 of 245 INFORMATION USE Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 2022 ft. 1.25 in. (Level 3) for  60 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
All Definition(s):
None Basis: Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). For Callaway Plant SFP Level 3 has been set at plant elevation 2022 ft. 1.25 in. (~11 in. above the top of the spent fuel racks) as indicated on EC-LI-0059B in the Control Room or EC-LI-0059A in the Auxiliary Building MG Set Room. Backup indication is also available on EC-LI-0060A in the Auxiliary Building MG Set Room.
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. 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 AG2
 
Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 62 of 245 INFORMATION USE Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas:  Control Room (SD-RE-33)  OR  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 preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
Callaway Basis Reference(s):
: 1. FSAR Section 12.3 Table 12.3-2 Area Radiation Monitors 2. NEI 99-01 AA3 Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 63 of 245 INFORMATION USE
 
Category E - Independent Spent Fuel Storage Installation (ISFSI) EAL Group: Any (EALs in this category are applicable to any plant condition, hot or cold.) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.
An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated. 
 
ATTACHMENT 1 EAL Bases  Page 64 of 245 INFORMATION USE Category: ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than EITHER of the following:
* 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 Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the Callaway ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC). Basis: Overpacks are the HI-STORM UMAX VVM casks which 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 overpack does not include the transfer cask (ref. 1).
The value s shown represents 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification section 5.3.4 for radiation external to either a loaded MPC overpack 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 fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme  + Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 1 EAL Bases  Page 65 of 245 INFORMATION USE damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSIs are covered under ICs HU1 and HA1. Callaway Basis Reference(s):
: 1. Certificate of Compliance No. 1040 Appendix A Technical Specifications for the HI-STORM UMAX Canister Storage System 2. NEI 99-01 E-HU1
 
VCSNS ATTACHMENT 1 EAL Bases  Page 66 of 245 INFORMATION USE Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature  200&#xba;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 occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, D - Defueled).
The events of this category pertain to the following subcategories: 1. 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 emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16KV 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 electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
ATTACHMENT 1 EAL Bases  Page 67 of 245 INFORMATION USE
: 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.
 
ATTACHMENT 1 EAL Bases  Page 68 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:
CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: With the plant in Cold Shutdown, RCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1). However, if RCS level is being controlled below the pressurizer low level setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the 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 LCO 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 (or the lower limit of a level band), or a loss of the ability to monitor RCS  level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease 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 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 level, specified for the current plant conditions, Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 69 of 245 INFORMATION USE cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
Callaway Basis Reference(s):
: 1. FR-I.2 Response to Low Pressurizer Level
: 2. OTN-BB-00002 Reactor Coolant System Draining 
: 3. Technical Specification Section 3.9.7 Refueling Pool Water Level 4. NEI 99-01 CU1
 
ATTACHMENT 1 EAL Bases  Page 70 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: 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 any Table C-1 sump/tank level due to loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps / Tanks
* Containment Sumps
* Containment Normal Sumps
* Containment Instrument Sump
* PRT
* RCDT
* Auxiliary Building Sump Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.
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 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). Callaway ATTACHMENT 1 EAL Bases  Page 71 of 245 INFORMATION USE 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 (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Refueling evolutions that decrease 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-1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
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. NEI 99-01 CU1
 
NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 72 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory  EAL:
CA1.1 Alert Loss of RCS inventory as indicated by Reactor Vessel level < bottom of RCS hot leg ID (RVLIS Pumps Off < 73% or BBLI-53 A/B at 0 inches) Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis: When Reactor Vessel water level lowers 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.) Level monitoring instruments BB LI-53A/B and Computer Point BBL0053BB cannot sense level changes in the Reactor Vessel below the elevation of 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). The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2): Elevation of bottom of Reactor Vessel (ft) A1987.150Elevation of bottom ID of RCS hot leg penetration (ft) B2013.290Hot leg penetration (above vessel bottom) C = B - A (ft)26.140Height of vessel D (ft)41.245RVLIS indication corresponding to the top of the core: H = 100 x C / D (%)63.377RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81%---OCA at H (%)8.949Bottom ID of RCS loop, including channel uncertainties: H + OCA (%)72.327Rounded 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 occurred. 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 level of plant safety. For this EAL, a lowering of RCS water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up Callaway NEI 99-01 Basis:&#xb6;6.5%
ATTACHMENT 1 EAL Bases  Page 73 of 245 INFORMATION USE 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 CS1.
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 CA1
 
ATTACHMENT 1 EAL Bases  Page 74 of 245 INFORMATION USE 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 min. (Note 1) AND EITHER
* UNPLANNED increase in any Table C-1 Sump / Tank level
* Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-1 Sumps / Tanks
* Containment Sumps
* Containment Normal Sumps
* Containment Instrument Sump
* PRT
* RCDT
* Auxiliary Building Sump Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: In Cold Shutdown mode, the 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.
In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated Callaway ATTACHMENT 1 EAL Bases  Page 75 of 245 INFORMATION USE 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 precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, the inability to monitor 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 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 CS1.
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. NEI 99-01 CA1
 
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 76 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability  EAL:
CS1.1 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 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.
Basis: When Reactor Vessel water level lowers to 2012.79 ft (ref. 1), water level is six inches below the elevation of the bottom of the RCS hot leg penetration. When Reactor Vessel water level drops significantly below the elevation of 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 elevation of the bottom of the RCS hot leg penetration can be monitored only by RVLIS. Level monitoring instruments BB LI-53A/B and Computer Point BBL0053BB cannot sense level changes in the Reactor Vessel below the elevation of 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). The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2): Elevation of bottom of Reactor Vessel (ft) A1987.150Elevation of bottom ID of RCS hot leg penetration (ft) B2013.290Six inches below hot leg penetration (above vessel bottom) C = B - A - 0.5 (ft)25.640Height of vessel D (ft)41.245RVLIS indication corresponding to the top of the core: H = 100 x C / D (%)62.165RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81%---OCA at H (%)8.937Bottom ID of RCS loop, including channel uncertainties: H + OCA (%)71.102Rounded upward to nearest 1% (RVLIS range is 0 - 120% in 2% increments)72 Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a 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 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 Callaway ATTACHMENT 1 EAL Bases  Page 77 of 245 INFORMATION USE 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, 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 CS1.1 and CS2.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1 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. OSP-GT-00003 Containment Closure 
: 5. NEI 99-01 CS1 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 78 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability  EAL:
CS1.2 Site Area Emergency With CONTAINMENT CLOSURE established, RVLIS Pumps Off < 65% (Top of Fuel) Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to Callaway, Containment Closure is established when the requirements of OSP-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 Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2):
Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a 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 Callaway ATTACHMENT 1 EAL Bases  Page 79 of 245 INFORMATION USE 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, 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 CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1 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. OSP-GT-00003 Containment Closure
: 5. NEI 99-01 CS1 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 80 of 245 INFORMATION USE 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 be monitored for  30 min. (Note 1)  AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude to indicate core uncovery
* Manipulator crane radiation monitor SD-RE-41 > 10,000 mR/hr
* Erratic Source Range Monitor indication Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-1 Sumps / Tanks
* Containment Sumps
* Containment Normal Sumps
* Containment Instrument Sump
* PRT
* RCDT
* Auxiliary Building Sump Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: In Cold Shutdown mode, the 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-1). <#>Visual observation of UNISOLABLE RCS leakage &#xb6;Callaway ATTACHMENT 1 EAL Bases  Page 81 of 245 INFORMATION USE 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).
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 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, fuel damage is probable.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor 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 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. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1 Callaway Basis Reference(s):
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 82 of 245 INFORMATION USE
: 1. OTO-BB-00003-R014 Excess RCS Leakage
: 2. OSP-BB-00009 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. Calc. 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-1 9. NEI 99-01 CS1
 
ATTACHMENT 1 EAL Bases  Page 83 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged  EAL:
CG1.1 General Emergency RVLIS Pumps Off < 65% (Top of Fuel) for  30 min. (Note 1)  AND Any Containment Challenge indication, Table C-2 Note  1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note  6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.
Table C-2 Containment Challenge Indications
* CONTAINMENT CLOSURE not established (Note 6)
* 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 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.
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 Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2): Callaway ATTACHMENT 1 EAL Bases  Page 84 of 245 INFORMATION USE Three conditions are associated with a challenge to Containment integrity: 1. CONTAINMENT COSURE 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. 4). If containment 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% 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 AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and ERFIS (ref. 6, 7). The hydrogen monitors require a 2 hour 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. If an actual hydrogen concentration measurement is unavailable, CA-3 may be used to estimate the Containment atmosphere hydrogen concentration (ref. 10).
ATTACHMENT 1 EAL Bases  Page 85 of 245 INFORMATION USE This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If 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 required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor 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 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. 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. 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 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 86 of 245 INFORMATION USE
: 4. OSP-GT-00003 Containment Closure
: 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. Calc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102 10. NEI 99-01 CS1  10.CA-3 Hydrogen Flammability in Containment&#xb6; 11 ATTACHMENT 1 EAL Bases  Page 87 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction 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 be monitored for  30 min. (Note 1)  AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude to indicate core uncovery
* Manipulator crane radiation monitor SD-RE-41 > 10,000 mR/hr
* Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-2 Note  1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note  6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-1 Sumps / 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:
<#>Visual observation of UNISOLABLE RCS leakage &#xb6; ATTACHMENT 1 EAL Bases  Page 88 of 245 INFORMATION USE 5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: 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-1). 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).
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 COSURE 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. 15). If containment closure is re-Callaway ATTACHMENT 1 EAL Bases  Page 89 of 245 INFORMATION USE established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur. 2. Containment hydrogen  4% - The 4% 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. 9). Two Containment hydrogen monitors (GS AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and ERFIS (ref. 10, 11). The hydrogen monitors require a 2 hour 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 Containment 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 IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If 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 required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various ). If an actual hydrogen concentration measurement is unavailable, CA-3 may be used to estimate the Containment atmosphere hydrogen concentration (ref. 14).NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 90 of 245 INFORMATION USE 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 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 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. 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. 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. FSAR Section 12.3.3.4
: 5. FSAR Table 12.3-2
: 6. Calc. 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-1 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. Calc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102 14. OSP-GT-00003 Containment Closure  15. NEI 99-01 CG1 14.CA-3 Hydrogen Flammability in Containment&#xb6; 1516 ATTACHMENT 1 EAL Bases  Page 91 of 245 INFORMATION USE 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 NB01 and NB02 reduced to a single power source for  15 min. (Note 1)    AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-3  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
ATTACHMENT 1 EAL Bases  Page 92 of 245 INFORMATION USE Basis: For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the emergency buses.
4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2, 3).
In addition, NB01 and NB02 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 XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes. This cold condition EAL is equivalent to the hot condition EAL SA1.1. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 93 of 245 INFORMATION USE 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):
: 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-R011 Main Step-Up Transformer Backfeed - IPTE 5. NEI 99-01 CU2
 
ATTACHMENT 1 EAL Bases  Page 94 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power 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 buses NB01 and NB02 for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-3  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis: For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. The emergency 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). Callaway ATTACHMENT 1 EAL Bases  Page 95 of 245 INFORMATION USE 4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2, 3).
In addition, NB01 and NB02 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 XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes. This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. 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 due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or RS1.
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-R011 Main Step-Up Transformer Backfeed - IPTE 5. NEI 99-01 CA2
 
NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 96 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:
CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200
&deg;F (Note 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):
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;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 MODE5). The most limiting temperature indication should be used. For example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (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. 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 due to loss of decay heat removal capability Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 97 of 245 INFORMATION USE inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Callaway Basis Reference(s):
: 1. Callaway Technical Specifications Table 1.1-1 
: 2. OTG-ZZ-00001 Plant Heatup 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 99-01 CU3
 
ATTACHMENT 1 EAL Bases  Page 98 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:
CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for  15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
None Basis: Reactor Vessel water level is normally monitored using the following instruments (ref. 2):
* RCS Loop level indications):
* Indicators on RL018: BB LI-53A, RCS (LOOP 1) HOT LEG LEV BB LI-53B, RCS (LOOP 4) HOT LEG LEV
* Computer points: BBL0053A, RCS LOOP 1 HOT LEG LEVEL BBL0053B, RCS LOOP 4 HOT LEG LEVEL BBL053BB, RCS LOOP LEVEL - CTMT VENTED
* RVLIS LI-1311, LI-1312, LI-1321, and LI-1322 (if in service) (ref. 3, 4)
* Visual observation (if vessel head is removed) (ref. 5) 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).
Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;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 MODE5). The most limiting temperature indication should be used. For Callaway ATTACHMENT 1 EAL Bases  Page 99 of 245 INFORMATION USE example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (ref. 6, 7, 8).
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 would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Callaway Basis Reference(s):
: 1. Callaway Technical Specifications Table 1.1-1 
: 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-00018 Draining the Refueling Pool
: 6. OTG-ZZ-00001 Plant Heatup Cold Shutdown to Hot Standby 7. OSP-BB-00007 RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2  8. FSAR Section 7.2.2.3.2
: 9. NEI 99-01 CU3
 
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 100 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED increase in RCS temperature to > 200
&deg;F for > Table C-4 duration  (Notes 1, 10) OR UNPLANNED RCS pressure increase > 10 psig (This EAL does not apply during water-solid plant conditions)  Note 1: The Emergency Coordinator should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. Note 10: Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal
. Table C-4: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Status Heat-up Duration Intact (but not REDUCED INVENTORY) N/A 60 min.* Not intact    OR REDUCED INVENTORY established 20 min.* not established 0 min.
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.
UNPLANNED -. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. due to a loss of RCS cooling ATTACHMENT 1 EAL Bases  Page 101 of 245 INFORMATION USE 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.).
 
ATTACHMENT 1 EAL Bases  Page 102 of 245 INFORMATION USE Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200&deg;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 MODE5). The most limiting temperature indication should be used. For example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (ref. 2, 3, 4).
RCS pressure instrument BB 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 Heat-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-up Duration Thresholds 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 reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The 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 would be via IC CS1 or RS1. Callaway Basis Reference(s):
: 1. Callaway Technical Specifications Table 1.1-1 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 103 of 245 INFORMATION USE
: 2. OTG-ZZ-00001 Plant Heatup 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. OTG-ZZ-00006 Plant Cooldown Hot Standby To Cold Shutdown 6. NEI 99-01 CA3
 
ATTACHMENT 1 EAL Bases  Page 104 of 245 INFORMATION USE 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 buses for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis: The purpose of this EAL is to recognize a loss 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. The vital DC buses are the following 125 VDC Class 1E buses (ref. 1): Division 1:
* NK01
* NK03 Division 2:
* NK02
* NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement 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.
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 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 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, NK02 and NK03. For this reason, batteries NK11 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. 2, 3, 4). Callaway ATTACHMENT 1 EAL Bases  Page 105 of 245 INFORMATION USE Minimum DC bus voltage is 107.0 VDC (ref. 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):
* NK EI-1 (NK01)
* NK EI-2 (NK02)
* NK EI-3 (NK03)
* NK EI-4 (NK04) This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1. This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.
As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.
Callaway Basis Reference(s):
: 1. E-21010(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-1 III.B
: 6. ECA-0.0 Loss of All AC Power
: 7. NEI 99-01 CU4 NEI 99-01 Basis
&#xb6; ATTACHMENT 1 EAL Bases  Page 106 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 ORO communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods System Onsite ORO NRC Gaitronics X  Plant Radios X  Plant Emergency Dedicated Phones X  Plant Telephone System X X X ENS (Red Phone) Line  X X Back-Up Radio System  X  Sentry Notification System  X 
 
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Definition(s):
None Basis: Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1, 2). Callaway ATTACHMENT 1 EAL Bases  Page 107 of 245 INFORMATION USE
: 1. Gai-tronics system The Gai-tronics 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 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 fuel 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 800 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 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 are located throughout the power block, in the main control room, in the various buildings 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 public 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.
ATTACHMENT 1 EAL Bases  Page 108 of 245 INFORMATION USE
: 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 back up notification of offsite agencies and the coordination of offsite activities during a radio logic a1 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 warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Callaway, Gasconade, Montgomery and Osage County EOCs 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. NEI 99-01 CU5 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 109 of 245 INFORMATION USE Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
CA6.1 Alert The occurrence of any Table C-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 Table C-6 Hazardous Events  Seismic event (earthquake)  Internal or external FLOODING event  High winds or tornado strike  FIRE  EXPLOSION  Other events with similar hazard characteristics as determined by the 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 overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
ATTACHMENT 1 EAL Bases  Page 110 of 245 INFORMATION USE FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a component or structure 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 component or structure. Basis:
* Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-00001, 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).
* 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 a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Callaway <#>The significance of seismic events are discussed under EAL HU2.1 (ref.
1).&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 111 of 245 INFORMATION USE The first conditional 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.
The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS1 or RS1. Callaway Basis Reference(s):
: 1. OTO-SG-00001 Seismic Event
: 2. IPE Section 3.4.2.3 Results of the Vulnerability Screening 3. UFSAR Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations 4. UFSAR Section 3.3.1.1 Design Wind Loadings
: 5. UFSAR Appendix 9.5B Fire Hazard Analysis and Site Combustion Restrictions  6. NEI 99-01 CA6
 
ATTACHMENT 1 EAL Bases  Page 112 of 245 INFORMATION USE Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. 1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. 3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
: 4. Fire  Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown
: 5. 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.
: 6. Emergency Coordinator Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based ATTACHMENT 1 EAL Bases  Page 113 of 245 INFORMATION USE 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.
 
ATTACHMENT 1 EAL Bases  Page 114 of 245 INFORMATION USE Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:
HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift 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):
SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.
HOSTILE ACTION - An act toward 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).
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).
Callaway NEI 99-01 Basis:
ATTACHMENT 1 EAL Bases  Page 115 of 245 INFORMATION USE 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 &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan . The first threshold references the  Shift Security Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR &sect; 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 (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the Callaway Plant Security Plan and DBT (ref. 1).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan and DBT (ref. 1). Escalation of the emergency classification level would be via IC HA1.
Callaway Basis Reference(s):
: 1. Callaway Plant Security Plan and DBT (Safeguards) 2. EIP-ZZ-SK001, Response to Security Threat
: 3. SDP-CP-00003 Security Contingency Events
: 4. OTO-SK-00002 - Plant Security Event - Aircraft Threat  5. NEI 99-01 HU1
 
ATTACHMENT 1 EAL Bases  Page 116 of 245 INFORMATION USE Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.1 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 min. of the site 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 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).
OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis: 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 NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 117 of 245 INFORMATION USE This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security procedures. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.
In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan and DBT (ref. 1). Callaway Basis Reference(s):
: 1. Callaway Plant Security Plan and DBT (Safeguards) 2. EIP-ZZ-SK001, Response to Security Threat
: 3. SDP-CP-00003 Security Contingency Events
: 4. OTO-SK-00002 - Plant Security Event - Aircraft Threat  5. NEI 99-01 HA1
 
ATTACHMENT 1 EAL Bases  Page 118 of 245 INFORMATION USE Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:
HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor  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 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 around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding 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 . 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 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 119 of 245 INFORMATION USE (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan and DBT (ref. 1). Escalation of the emergency classification level would be via IC HG1. Callaway Basis Reference(s):
: 1. Callaway Plant Security Plan and DBT (Safeguards) 2. EIP-ZZ-SK001, Response to Security Threat
: 3. SDP-CP-00003 Security Contingency Events
: 4. OTO-SK-00001 - Plant Security Event - Hostile Intrusion 5. OTO-SK-00002 - Plant Security Event - Aircraft Threat  6. NEI 99-01 HS1
 
ATTACHMENT 1 EAL Bases  Page 120 of 245 INFORMATION USE Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility EAL:
HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor AND EITHER of the following has occurred:
* Any of the following safety functions cannot be controlled or maintained - Reactivity control
- Core cooling
- RCS heat removal OR
* Damage to spent fuel has occurred or is IMMINENT 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 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).
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  PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area.
Basis: The security shift supervision is defined as the Security Shift Supervisor. This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 121 of 245 INFORMATION USE fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. 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). Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan & DBT (ref.1). Callaway Basis Reference(s):
: 1. Callaway Plant Security Plan and DBT (Safeguards) 2. EIP-ZZ-SK001, Response to Security Threat
: 3. SDP-CP-00003 Security Contingency Events
: 4. OTO-SK-00001 - Plant Security Event - Hostile Intrusion 5. OTO-SK-00002 - Plant Security Event - Aircraft Threat 6. NEI 99-01 HG1
 
ATTACHMENT 1 EAL Bases  Page 122 of 245 INFORMATION USE 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: Ground motion acceleration of 0.12g horizontal or vertical is the Operating Basis Earthquake for Callaway (ref. 1).
Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold (ref. 2).
When the seismic recorder indicates that the OBE has been exceeded, 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.
OTO-SG-00001, Seismic Event provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 2) To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Callaway. Alternatively, near real-time seismic activity can be accessed via the NEIC website: http://earthquake.usgs.gov/eqcenter/  per OTO-SG-00001,Callaway ATTACHMENT 1 EAL Bases  Page 123 of 245 INFORMATION USE This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.
Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Callaway Basis Reference(s):
: 1. FSAR Section 3.8.6.4.1 Design and Analysis Procedures 2. OTO-SG-00001, Seismic Event
: 3. NEI 99-01 HU2 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 124 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:
HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):
PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-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 EAL CA6.1 or SA9.1.
A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
Callaway Basis Reference(s):
: 1. OTO-ZZ-00012 Severe Weather 
: 2. NEI 99-01 HU3 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 125 of 245 INFORMATION USE 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 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: 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. Callaway The internal flooding areas of concern are the Control Building, Battery Room and ESF Switchgear Room (ref.1). NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 126 of 245 INFORMATION USE Callaway Basis Reference(s):
: 1. IPE Section 3.4.2.3 Results of the Vulnerability Screening 2. NEI 99-01 HU3
 
ATTACHMENT 1 EAL Bases  Page 127 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:
HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)
 
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area.
Basis: As used here, the term "offsite" is meant to be areas external to the Callaway PROTECTED AREA.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
Callaway Basis Reference(s):
: 1. NEI 99-01 HU3
 
Callaway NEI 99-01 Basis:&#xb6; ATTACHMENT 1 EAL Bases  Page 128 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:
HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability:
All Definition(s):
None Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
Callaway Basis Reference(s):
: 1. NEI 99-01 HU3
 
Callaway None&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 129 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):  Report from the field (i.e., visual observation)  Receipt of multiple (more than 1) fire alarms or indications  Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table H-1  Fire Areas
* Area 5
* Containment
* Aux Feed Pump Rooms
* Auxiliary Building
* Diesel Generator Building
* UHS Cooling Tower
* ESW Pumphouse
* Control Building/Communications Corridor
* RWST
* Fuel Building 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. 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 field reports must be made within the 15 minute Callaway ATTACHMENT 1 EAL Bases  Page 130 of 245 INFORMATION USE time limit or a classification must be made. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of the fire detection alarm. Table H-1 Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.
Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. 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. NEI 99-01 HU4 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 131 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table H-1  Fire Areas
* Area 5
* Containment
* Aux Feed Pump Rooms
* Auxiliary Building
* Diesel Generator Building
* UHS Cooling Tower
* ESW Pumphouse
* Control Building/Communications Corridor
* RWST
* Fuel Building 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. Basis: The 30 minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1. Callaway ATTACHMENT 1 EAL Bases  Page 132 of 245 INFORMATION USE Table H-1 Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in 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. NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 133 of 245 INFORMATION USE Callaway Basis Reference(s):
: 1. FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown  2. NEI 99-01 HU4
 
ATTACHMENT 1 EAL Bases  Page 134 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 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.
In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Callaway Basis Reference(s):
: 1. NEI 99-01 HU4 Callaway None&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 135 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 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 offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Callaway Basis Reference(s):
: 1. NEI 99-01 HU4
 
Callaway None&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 136 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA5.1 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 inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. OTO-ZZ-00001 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). Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.
Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. 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 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 137 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:
HS5.1 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 min. (Note 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: 1 - 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 starts when the last licensed operator leaves the Control Room. The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. OTO-ZZ-00001 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 fifteen minute time for transfer starts when the Control Room begins to be evacuated (not when OTO-ZZ-00001 is entered). 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.
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 manner. Primary All&#xb6;Callaway ATTACHMENT 1 EAL Bases  Page 138 of 245 INFORMATION USE 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 shutdown 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 is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.
The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Escalation of the emergency classification level would be via IC FG1 or CG1 Callaway Basis Reference(s):
: 1. OTO-ZZ-00001 Control Room Inaccessibility
: 2. OTS-ZZ-00001 Cooldown from Outside the Control Room 3. NEI 99-01 HS6
 
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 139 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:
HU6.1  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. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (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 warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Unusual Event. Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 140 of 245 INFORMATION USE 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 HU7
 
ATTACHMENT 1 EAL Bases  Page 141 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:
HA6.1 Alert Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward 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).
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 Operations 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 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. 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 Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 142 of 245 INFORMATION USE
: 3. NEI 99-01 HA7
 
ATTACHMENT 1 EAL Bases  Page 143 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:
HS6.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward 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)
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 Operations 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 warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a Site Area Emergency.
Callaway Basis Reference(s): Callaway NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 144 of 245 INFORMATION USE
: 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 HS7
 
ATTACHMENT 1 EAL Bases  Page 145 of 245 INFORMATION USE Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:
HG6.1  General Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward 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).
IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures (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. Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.
Callaway ATTACHMENT 1 EAL Bases  Page 146 of 245 INFORMATION USE 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.
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 HG7 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 147 of 245 INFORMATION USE Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200&#xba;F); EALs in this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories: 1. Loss of Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 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 warrant emergency classification. Losses of indicators are in this subcategory.
: 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
: 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
: 6. RTS 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 referred to as ATTACHMENT 1 EAL Bases  Page 148 of 245 INFORMATION USE 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 cause a threat to fuel clad, RCS and containment integrity. 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. 8. 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 Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory. 
 
ATTACHMENT 1 EAL Bases  Page 149 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:
SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: Basis: The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3). Callaway ATTACHMENT 1 EAL Bases  Page 150 of 245 INFORMATION USE In addition, NB01 and NB02 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).
Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are 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. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level would be via IC SA1. 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-R011 Main Step-Up Transformer Backfeed - IPTE 5. NEI 99-01 SU1
 
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.&#xb6; NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 151 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:
SA1.1 Alert AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 reduced to a single power source for  15 min. (Note 1)    AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis:
ATTACHMENT 1 EAL Bases  Page 152 of 245 INFORMATION USE For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3). In addition, NB01 and NB02 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).
Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.
This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator). A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator. A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level would be via IC SS1. Callaway Basis Reference(s):
: 1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram) Callaway Basis:
&#xb6;The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 153 of 245 INFORMATION USE
: 2. FSAR Site Addenda Section 8.2.1.2
: 3. FSAR Section 8.3.1
: 4. OTS-MA-00001-R011 Main Step-Up Transformer Backfeed - IPTE 5. NEI 99-01 SA1
 
ATTACHMENT 1 EAL Bases  Page 154 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:
SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3). Callaway ATTACHMENT 1 EAL Bases  Page 155 of 245 INFORMATION USE In addition, NB01 and NB02 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).
Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes (ref. 5).
The interval begins when both offsite and onsite AC power capability are lost. This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.
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-R011 Main Step-Up Transformer Backfeed - IPTE 5. ECA-0.0 Loss of All AC Power
: 6. NEI 99-01 SS1
 
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 156 of 245 INFORMATION USE 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 capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 AND EITHER:
* Restoration of at least one emergency bus in < 4 hours is not likely (Note 1)
* CSFST Core Cooling RED Path conditions met Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
1 - 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 NB01 and NB02 either for greater then the Callaway Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1, 2) or that has resulted in indications of an 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). For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. Callaway ATTACHMENT 1 EAL Bases  Page 157 of 245 INFORMATION USE The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 4, 5 6). In addition, NB01 and NB02 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 XMA01 and unit auxiliary transformer XMR02. 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). Additional sources of offsite power are available from diesel generators such as the Alternate Emergency Power Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes (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 barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 3). Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200&#xba;F or core exit T/Cs are reading greater than or equal to 706&#xba;F with RCS subcooling less than or equal to 30&#xba;F [50&#xba;F], and RVLIS indication is less than or equal to that specified based on the number of RCPs running (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, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC 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. NEI-9901 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 158 of 245 INFORMATION USE 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 of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
Callaway Basis Reference(s):
: 1. FSAR Section 8.3A.5
: 2. BO-01 Station Blackout (SBO) Coping Duration, sh 1 3. CSF-1 Critical Safety Function Status Trees (CSFST) Figure 2, Core Cooling 4. E-21001(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-00001-R011 Main Step-Up Transformer Backfeed - IPTE 8. ECA-0.0 Loss of All AC Power
: 9. NEI 99-01 SG1
 
ATTACHMENT 1 EAL Bases  Page 159 of 245 INFORMATION USE Category: S -System Malfunction Subcategory: 1 - 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 capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 for  15 min. AND Loss of all 125 VDC power based on battery bus voltage indications < 107 VDC on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-1  AC Power Sources Offsite:
* Safeguards XFMR A or B via ESF LTC XFMR XNB01
* Startup XFMR XMR01 via ESF LTC XFMR XNB02
* Main XFMR XMA01 backfed via UAT XFMR XMA02 (only if already aligned) Onsite:
* EDG NE01
* EDG NE02  Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4.16KV emergency buses NB01 and NB02 for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.
For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1). Callaway ATTACHMENT 1 EAL Bases  Page 160 of 245 INFORMATION USE NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 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 XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3). In addition, NB01 and NB02 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).
Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMR02. 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 Supply (AEPS) or portable generation sources. Credit can be taken for these sources if they are capable of carrying an NB bus and are aligned within 15 minutes (ref. 5).
The vital DC buses are the following 125 VDC Class 1E buses (ref. 6): Division 1:
* NK01
* NK03 Division 2:
* NK02
* NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement 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.
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 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 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, NK02 and NK03. For this reason, batteries NK11 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):
* NK EI-1 (NK01)
* NK EI-2 (NK02)
* NK EI-3 (NK03)
* NK EI-4 (NK04) This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, NEI-9901 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 161 of 245 INFORMATION USE containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both 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.
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-R011 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
: 8. FSAR Section 8.3.2
: 9. Calculation NK-10, NK System DC Voltage Drop
: 10. FSAR Table 8.3A-1 III.B
: 11. NEI 99-01 SG8
 
ATTACHMENT 1 EAL Bases  Page 162 of 245 INFORMATION USE 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 NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: The vital DC buses are the following 125 VDC Class 1E buses (ref. 1): Division 1:
* NK01
* NK03 Division 2:
* NK02
* NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement 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.
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 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 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, NK02 and NK03. For this reason, batteries NK11 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. 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 EI-1 (NK01)
* NK EI-2 (NK02)
* NK EI-3 (NK03)
* NK EI-4 (NK04)  Callaway ATTACHMENT 1 EAL Bases  Page 163 of 245 INFORMATION USE This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1. Callaway Basis Reference(s):
: 1. E-21010(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-1 III.B
: 6. ECA-0.0 Loss of All AC Power
: 7. NEI 99-01 SS8 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 164 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table S-2 Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* Core Exit T/C temperature
* Level in at least one S/G
* Auxiliary or emergency feed flow in at least one S/G Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).
Callaway NEI 99-01 Basis:
ATTACHMENT 1 EAL Bases  Page 165 of 245 INFORMATION USE This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and 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 IC SA3. Callaway Basis Reference(s):
: 1. UFSAR Section 7.5 Safety-Related Display Instrumentation 2. OTO-RJ-00001 Loss of Plant Computer
: 3. NEI 99-01 SU2
 
ATTACHMENT 1 EAL Bases  Page 166 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1) AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table S-2 Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* Core Exit T/C temperature
* Level in at least one S/G
* Auxiliary or emergency feed flow in at least one S/G Table S-3 Significant Transients
* Reactor trip
* Runback  25% thermal power
* Electrical load rejection > 25% electrical load
* ECCS actuation Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis: Callaway ATTACHMENT 1 EAL Bases  Page 167 of 245 INFORMATION USE SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).
Significant transients are listed in Table S-2 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than or equal to 25% thermal power change, electrical load rejections of greater than 25% full electrical load or ECCS (SI) injection actuations.
This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
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 FS1 or IC RS1 Callaway Basis Reference(s):
: 1. UFSAR Section 7.5 Safety-Related Display Instrumentation 2. OTO-RJ-00001 Loss of Plant Computer
: 3. NEI 99-01 SA2 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 168 of 245 INFORMATION USE
 
ATTACHMENT 1 EAL Bases  Page 169 of 245 INFORMATION USE 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 Section 3.4.16 limits Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: The specific iodine activity is limited to either  60 &#xb5;Ci/gm Dose Equivalent I-131 or  1.0 &#xb5;Ci/gm Dose Equivalent I-131 for a > 48 hr continuous period. The specific Xe-133 activity is limited to  225 &#xb5;Ci/gm Dose Equivalent Xe-133 (ref 1, 2). This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.
Callaway Basis Reference(s):
: 1. Callaway Technical Specifications section 3.4.16 RCS Specific Activity  2. OTO-BB-00005 High Coolant Activity
: 3. NEI 99-01 SU3
 
Callaway XENEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 170 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:
SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for  15 min. OR RCS identified leakage > 25 gpm for  15 min. OR Leakage from the RCS to a location outside containment > 25 gpm for  15 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage. The Personal 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, 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 injection or 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 pressure boundary leakage, or
* RCS leakage through a steam generator to the secondary system (ref. 2). Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 2).
Pressure Boundary leakage is leakage (except SG leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2) 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 Callaway ATTACHMENT 1 EAL Bases  Page 171 of 245 INFORMATION USE 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 FA1.1.
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 calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
Callaway Basis Reference(s):
: 1. OSP-BB-00009 RCS Inventory Balance 
: 2. Callaway Technical Specifications Definitions section 1.1 3. UFSAR Section 5.2.5.2.1 Intersystem Leakage
: 4. OTO-BB-00003-R014 Excess RCS Leakage
: 5. NEI 99-01 SU4
 
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 172 of 245 INFORMATION USE Category: S - System Malfunction 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 successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability:
1 - Power Operation Definition(s):
None Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation 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. 1).
Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RTS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4). For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB-HS-1 on Panel RL003 or SB-HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as opening PG19 and PG20 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4). Callaway ATTACHMENT 1 EAL Bases  Page 173 of 245 INFORMATION USE Following any automatic RTS trip signal, E-0 (ref. 2) and /FR-S.1 (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 manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event (ref. 4). A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates 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 SA6.1. If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions. If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.
This IC addresses a failure of the 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 shutdown the reactor (e.g., initiate a manual reactor trip ). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip ) using a different switch). Depending upon several factors, the initial or subsequent effort to manually 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 plant's 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 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 174 of 245 INFORMATION USE injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip  will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., 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 shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted. 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 Function Status Trees - Subcriticality 4. FR-S.1 Response to Nuclear Power Generation/ATWS
: 5. FSAR Section 7.7.1 6 NEI 99-01 SU5
 
ATTACHMENT 1 EAL Bases  Page 175 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:
SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power  5% after any manual trip action was initiated  AND A subsequent automatic trip or manual trip action taken at the reactor control console  (SB-HS-1 or SB-HS-42) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability:
1 - Power Operation Definition(s):
None Basis: This EAL addresses 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 down the reactor (reactor power < 5%). (ref. 1). Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from 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). For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB-HS-1 on Panel RL003 or SB-HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as opening PG19 and PG20 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4). Callaway ATTACHMENT 1 EAL Bases  Page 176 of 245 INFORMATION USE Following the failure of any manual trip signal, E-0 (ref. 2) and FR-S.1 (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 full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an Unusual Event (ref. 4).
A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates auxiliary 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 shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually 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 plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip  will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 177 of 245 INFORMATION USE 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 RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted. 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 Function Status Trees - Subcriticality 4. FR-S.1 Response to Nuclear Power Generation/ATWS
: 5. FSAR Section 7.7.1
: 6. NEI 99-01 SU5
 
ATTACHMENT 1 EAL Bases  Page 178 of 245 INFORMATION USE 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-HS-1 or SB-HS-42) are not successful in shutting down the reactor as indicated by reactor power  5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability:
1 - Power Operation Definition(s):
None Basis: This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor (reactor power < 5%) followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (ref. 1).
For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB-HS-1 on Panel RL003 or SB-HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as opening PG19 and PG20 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).
5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent Callaway ATTACHMENT 1 EAL Bases  Page 179 of 245 INFORMATION USE subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 3, 4).
Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RTS.
A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".
The plant response to the failure of an automatic or manual reactor trip  will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
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 Function Status Trees - Subcriticality 4. FR-S.1 Response to Nuclear Power Generation/ATWS
: 5. FSAR Section 7.7.1
: 6. NEI 99-01 SA5 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 180 of 245 INFORMATION USE 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 Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power  5%  AND All actions to shut down the reactor are not successful as indicated by reactor power  5% AND EITHER
:
* CSFST Core Cooling RED Path conditions met
* CSFST Heat Sink RED Path conditions met  Mode Applicability:
1 - Power Operation Definition(s):
None Basis: This EAL addresses the following:
* Any automatic reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1), and
* Indications that either core cooling is extremely challenged or heat removal is extremely challenged. The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.
Reactor shutdown achieved by use of FR-S.1 Response to Nuclear Power Generation/ATWS (such as opening PG19 and PG20 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 4). 5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a Callaway ATTACHMENT 1 EAL Bases  Page 181 of 245 INFORMATION USE normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1, 4).
Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2). Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200&#xba;F or core exit T/Cs are reading greater than or equal to 706&#xba;F with Reactor Vessel Lower Range level less than or equal to that specified based on the number of RCPs running (ref. 2). Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 2). Specifically, Heat Sink RED PATH conditions exist if narrow range level in at least on steam generator is not greater than or equal to 7% (25% ACC) and total feedwater flow to the steam generators is less than or equal to 285,000 lbm/hr. (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 emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG1 or FG1.
Callaway Basis Reference(s):
: 1. CSF-1 Critical Safety Function Status Trees - Figure 1 Subcriticality 2. CSF-1 Critical Safety Function Status Tress - Figure 2 Core Cooling 3. CSF-1 Critical Safety Function Status Tress - Figure 3 Heat Sink 4. FR-S.1 Response to Nuclear Power Generation/ATWS
: 5. NEI 99-01 SS5
 
NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 182 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 ORO communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite ORO NRC Gaitronics X  Plant Radios X  Plant Emergency Dedicated Phones X  Plant Telephone System X X X ENS (Red Phone) Line  X X Back-Up Radio System  X  Sentry Notification System  X 
 
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1, 2). Callaway C-5 ATTACHMENT 1 EAL Bases  Page 183 of 245 INFORMATION USE
: 1. Gai-tronics system The Gai-tronics 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 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 fuel 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 800 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 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 are located throughout the power block, in the main control room, in the various buildings 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 public 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.
ATTACHMENT 1 EAL Bases  Page 184 of 245 INFORMATION USE
: 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 back up notification of offsite agencies and the coordination of offsite activities during a radio logic a1 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 hot condition equivalent of the cold condition EAL CU5.1. This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Callaway, Gasconade, Montgomery and Osage County EOCs 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. NEI 99-01 SU6 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 185 of 245 INFORMATION USE Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control. EAL:
SU8.1 Unusual Event Any penetration is not isolated within 15 min. of a VALID containment isolation signal  OR Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for  15 min. (Note 9) (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. Note 9: One Containment Spray System train and one Containment Cooling System train comprise one full train of depressurization equipment. Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis: 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 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 fans are designed to start automatically in slow speed if not already running (ref. 3). Callaway ATTACHMENT 1 EAL Bases  Page 186 of 245 INFORMATION USE 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 EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible. The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner. This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers. 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. NEI 99-01 SU7 NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 187 of 245 INFORMATION USE 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 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 Table S-5 Hazardous Events  Seismic event (earthquake)  Internal or external FLOODING event  High winds or tornado strike  FIRE  EXPLOSION  Other events with similar hazard characteristics as determined by the Emergency Coordinator Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems ATTACHMENT 1 EAL Bases  Page 188 of 245 INFORMATION USE classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;  (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a component or structure 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 component or structure. Basis:
* Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-00001, 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).
* 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 a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.
The first condition 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. Callaway <#>The significance of seismic events are discussed under EAL HU2.1 (ref.
1).&#xb6;NEI 99-01 Basis:
&#xb6; ATTACHMENT 1 EAL Bases  Page 189 of 245 INFORMATION USE The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS1 or RS1. Callaway Basis Reference(s):
: 1. OTO-SG-00001 Seismic Event
: 2. IPE Section 3.4.2.3 Results of the Vulnerability Screening 3. UFSAR Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations 4. UFSAR Section 3.3.1.1 Design Wind Loadings
: 5. UFSAR Appendix 9.5B Fire Hazard Analysis and Site Combustion Restrictions  6. NEI 99-01 SA9 ATTACHMENT 1 EAL Bases  Page 190 of 245 INFORMATION USE Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200&#xba;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 products to the environment. The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System (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 includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
ATTACHMENT 1 EAL Bases  Page 191 of 245 INFORMATION USE
* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.
* The fission product barrier thresholds specified within a scheme reflect plant-specific Callaway design and operating characteristics.
* As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the 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 barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.
 
ATTACHMENT 1 EAL Bases  Page 192 of 245 INFORMATION USE Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:
FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-1) Mode Applicability:
1 - 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-1 (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 barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 Callaway Basis Reference(s):
: 1. NEI 99-01 FA1
 
Callaway NEI 99-01 Basis:
&#xb6;None&#xb6; ATTACHMENT 1 EAL Bases  Page 193 of 245 INFORMATION USE Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:
FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
* One barrier loss and a second barrier loss (i.e., loss - loss)
* One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
* One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.
Callaway Basis Reference(s): 1. NEI 99-01 FS1 Callaway NEI 99-01 Basis:
&#xb6;None&#xb6; ATTACHMENT 1 EAL Bases  Page 194 of 245 INFORMATION USE Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:
FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
* Loss of Fuel Clad, RCS and Containment barriers
* Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier
* Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier
* Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier Callaway Basis Reference(s):
: 1. NEI 99-01 FG1 Callaway NEI 99-01 Basis:
&#xb6;None&#xb6; ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases  Page 195 of 245 INFORMATION USE Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are: A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / RCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.1," the third Containment barrier Potential Loss in Category C would be assigned "CMT P-Loss C.3," etc.
If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category 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 barrier can occur. Barrier ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases  Page 196 of 245 INFORMATION USE Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,-, E.
ATTACHMENT 2 Page 197 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 197 of 245 Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or SG Tube Leakage None None 1. An automatic or manual ECCS (SI) actuation required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE 1. Operation of a standby charging pump is required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE 2. CSFST Integrity-RED Path conditions met 1. A leaking or RUPTURED SG is FAULTED outside of containment None B Inadequate Heat Removal 1. CSFST Core Cooling-RED Path conditions met 1. CSFST Core Cooling-ORANGE Path conditions met 2. CSFST Heat Sink-RED Path conditions met    AND  Heat sink is required None 1. CSFST Heat Sink-RED Path conditions met    AND  Heat sink is required None 1. CSFST Core Cooling-RED Path conditions met    AND  Restoration procedures not effective within 15 min. (Note 1) C CMT Radiation / RCS Activity 1. Containment radiation  > 2.80E+03 R/hr on GT-RE-59 (591) or GT-RE-60 (601) 2. Dose equivalent I-131 coolant activity > 300 &#xb5;Ci/gm 3. CVCS letdown radiation > 2.50E+01 &#xb5;Ci/ml on SJ-RE-01 (016) None 1. Containment radiation  > 6.40E+00 R/hr on GT-RE-59 (591) or GT-RE-60 (601)
None None 1. Containment radiation  > 8.06E+04 R/hr on GT-RE-59 (591) or GT-RE-60 (601)
D CMT Integrity or Bypass None None None None 1. Containment isolation is required AND EITHER:
* Containment integrity has been lost based on Emergency Coordinator judgment
* UNISOLABLE pathway from Containment to the environment exists 2. Indications of RCS leakage outside of Containment 1. CSFST Containment-RED Path conditions met 2. Containment hydrogen concentration  4% 3. Containment pressure > 27 psig with < one full train of depressurization equipment operating per design for 
> 15 min. (Note 1, 9) E EC Judgment 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the fuel clad barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the fuel clad barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier
 
ATTACHMENT 2 Page 198 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 198 of 245 Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:
None ATTACHMENT 2 Page 199 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 199 of 245 Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None ATTACHMENT 2 Page 200 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 200 of 245 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.
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  Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 201 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 201 of 245 Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 1. CSFST Core Cooling-ORANGE Path conditions met Definition(s):
None Basis: Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates  subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the 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  Plant-Specific&#xb6;indicatesGeneric&#xb6; ATTACHMENT 2 Page 202 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 202 of 245 Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 2. CSFST Heat Sink-RED Path conditions met    AND  Heat sink is required Definition(s):
None Basis: In combination with RCS Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the 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.1 is entered from CSFST Heat Sink-Red. Step 1 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&#xba;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 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.
Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 203 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 203 of 245 Callaway Basis Reference(s):
: 1. CSF-1 Critical Safety Function Status Trees Figure 3 Heat Sink 2. FR-H.1 Response to Loss of Secondary Heat Sink
: 3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B ATTACHMENT 2 Page 204 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 204 of 245 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:  1. Containment radiation > 2.80E+03 R/hr on GT-RE-59 (591) or GT-RE-60 (601) Definition(s):
None Basis: Containment radiation monitor readings greater than 2.8E+03 R/hr (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 &#xb5;Ci/cc dose equivalent I-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 5% clad failure depending on core inventory and RCS volume). This value is higher than that specified for RCS barrier Loss #3.
Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601). The threshold value of 2.8E+03 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 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 205 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 205 of 245 Callaway Basis Reference(s):
: 1. EPCI-9801 pg 27
: 2. OTA-SP-RM011 Radiation Monitor Control Panel RM-11 2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A ATTACHMENT 2 Page 206 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 206 of 245 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:  2. Dose equivalent I-131 coolant activity > 300 &#xb5;Ci/cc Definition(s):
None Basis: Dose Equivalent Iodine (DEI) is determined by Chemistry procedure CDP-ZZ-08100, 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 I-131 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 lost. (ref. 2).
This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation. Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 207 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 207 of 245 Callaway Basis Reference(s):
: 1. CDP-ZZ-08100 Post Accident Sampling Guidelines
: 2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B
 
ATTACHMENT 2 Page 208 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 208 of 245 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:  3. CVCS letdown radiation > 2.50E+01 &#xb5;Ci/ml 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 upstream 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-01 provides indication in the Control Room on Panel RM-11 channel 016 with a range of 1.7E-03 to 1.7E+03 &#xb5;Ci/ml (ref. 2, 3). The HI-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.
Callaway Basis Reference(s): 1. FSAR Section 9.3.4.2
: 2. FSAR Table 11.5-1
: 3. OTA-SP-RM011 Radiation Monitor Control Panel RM-11 4. HPCI-05-02 Gaseous and Liquid Radiation Monitor Setpoints Rev. 0, Note 11 5. NEI 99-01 Other Indications Fuel Clad Loss 5.A Plant-Specific&#xb6;Generic&#xb6;None&#xb6; ATTACHMENT 2 Page 209 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 209 of 245 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:  None 
 
ATTACHMENT 2 Page 210 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 210 of 245 Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold: None 
 
ATTACHMENT 2 Page 211 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 211 of 245 Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None 
 
ATTACHMENT 2 Page 212 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 212 of 245 Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold: 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within 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 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  Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 213 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 213 of 245 Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold: 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within 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 are to be used by the Emergency Coordinator  in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Callaway Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 214 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 214 of 245 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 < 1849 psig
* Steamline low pressure < 615 psig
* Containment high pressure > 3.5 psig 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 1.A will also be met. Callaway Basis Reference(s):
: 1. E-0 Reactor Trip or Safety Injection
: 2. E-3 Steam Generator Tube Rupture
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 215 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 215 of 245 Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold: 1. Operation of a standby charging pump is required by EITHER:
* UNISOLABLE 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 flow capacity of 130 gpm, and two centrifugal charging pumps each with a flow capacity of 150 gpm (ref. 1). Approximately 12 gpm of charging flow bypasses the RCS due to leakage through the RCP seals; thus, the Normal Charging Pump can deliver 130 gpm - 12 gpm = 118 gpm (rounded to 120 gpm for readability) (ref. 2). A second charging pump being required is indicative of a substantial RCS leak.
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) pump be placed in service 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 containment, 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 1.A will also be met.
Callaway Basis Reference(s):
: 1. UFSAR Table 9.3-9
: 2. UFSAR 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 1.A Generic&#xb6;Generic&#xb6; ATTACHMENT 2 Page 216 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 216 of 245
 
ATTACHMENT 2 Page 217 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 217 of 245 Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold: 2. CSFST Integrity-RED Path conditions met Definition(s):
None Basis:
The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity - RED path. CSFST RCS Integrity - Red Path plant conditions and associated PTS Limit Curve A indicates an extreme challenge to the safety function when plant parameters are to the left of the limit curve following excessive 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 causes rapid RCS cooldown while 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 1.B
 
Generic&#xb6;Generic&#xb6; ATTACHMENT 2 Page 218 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 218 of 245 Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold: None 
 
ATTACHMENT 2 Page 219 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 219 of 245 Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:  1. CSFST Heat Sink-RED path conditions met    AND  Heat sink is required Definition(s):
None Basis: In combination with FC Potential Loss B.2, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the 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.1 is entered from CSFST Heat Sink-Red. Step 1 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&#xba;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).
Generic 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 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. Plant-Specific&#xb6; ATTACHMENT 2 Page 220 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 220 of 245 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.
Callaway Basis Reference(s):
: 1. CSF-1 Critical Safety Function Status Trees Figure 3 Heat Sink 2. FR-H.1 Response to Loss of Secondary Heat Sink
: 3. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B
 
ATTACHMENT 2 Page 221 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 221 of 245 Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: 1. Containment radiation > 6.40E+00 R/hr on GT-RE-59 (591) or GT-RE-60 (601) Definition(s):
N/A Basis:
Containment radiation monitor readings greater than 6.4 E+00 R/hr (ref. 1) indicate the release of reactor coolant to the Containment. 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-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601). The threshold value of 6.4 E+00 R/hr is the HI (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.1 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):
: 1. EPCI-9801 pg 27
: 2. OTA-SP-RM011 Radiation Monitor Control Panel RM-11 3. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A  Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 222 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 222 of 245 Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None  B ATTACHMENT 2 Page 223 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 223 of 245 Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold: None ATTACHMENT 2 Page 224 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 224 of 245 Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None 
 
ATTACHMENT 2 Page 225 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 225 of 245 Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold: 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the 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 "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 RCS Barrier is lost.
Callaway Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A  Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 226 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 226 of 245 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 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 "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 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 cannot be monitored.
Callaway Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Plant-Specific&#xb6;Generic&#xb6;Director ATTACHMENT 2 Page 227 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 227 of 245 Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold: 1. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes. The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS 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) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment. Plant-Specific&#xb6;None.&#xb6; Generic&#xb6; ATTACHMENT 2 Page 228 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 228 of 245 Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs. The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED  Outside of Containment? P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of a standby charging (makeup) pump (RCS Barrier Potential Loss) Site Area Emergency per FS1.1 Alert per FA1.1 Requires an automatic or manual ECCS (SI) actuation (RCS Barrier Loss) Site Area Emergency per FS1.1 Alert per FA1.1
 
There is no Potential Loss threshold associated with RCS or SG Tube Leakage. Callaway Basis Reference(s):
: 1. E-2 Faulted Steam Generator Isolation
: 2. E-3 Steam Generator Tube Rupture
: 3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A ATTACHMENT 2 Page 229 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 229 of 245 Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold: None ATTACHMENT 2 Page 230 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 230 of 245 Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Loss Threshold: None 
 
ATTACHMENT 2 Page 231 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 231 of 245 Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:
: 1. CSFST Core Cooling-RED path conditions met    AND  Restoration procedures not effective within 15 min. (Note 1) Note 1: The Emergency Director 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. 1).
The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3). A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,200
&deg;F (ref. 1), Fuel Clad barrier is also lost.
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 Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Callaway Basis Reference(s):
: 1. CSF-1 Critical Safety Function Status Trees Figure 2 Core Cooling 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 Containment Potential Loss 2.A
 
Barrier: Containment Category: C. CMT Radiation/RCS Activity Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 232 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 232 of 245 Degradation Threat: Loss Threshold: None ATTACHMENT 2 Page 233 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 233 of 245 Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: 1. Containment radiation > 8.06E+04 R/hr on GT-RE-59 (591) or GT-RE-60 (601) Definition(s):
None Basis: Containment radiation monitor readings greater than 8.06 E+04 R/hr (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.
The readings are higher than that specified for Fuel Clad barrier Loss C.1 and RCS barrier 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 product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601) (ref. 2).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.
Callaway Basis Reference(s):
: 1. EPCI-9801 pg 27
: 2. OTA-SP-RM011 Radiation Monitor Control Panel RM-11 3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 234 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 234 of 245 Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold: 1. Containment isolation is required AND EITHER:
* Containment integrity has been lost based on Emergency Coordinator judgment
* UNISOLABLE pathway from containment to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds. First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.). Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
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 Plant-Specific&#xb6;None&#xb6; Generic&#xb6; ATTACHMENT 2 Page 235 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 235 of 245 or potential loss of containment but should be evaluated using the Recognition Category R ICs.
Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Leakage between two interfacing liquid systems, by itself, does not meet this threshold. Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
Callaway Basis Reference(s):
: 1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A 
 
ATTACHMENT 2 Page 236 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 236 of 245 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 Outside 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
* PZR/RCS Loop sample lines Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment. Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well. 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.1 to be met.
Callaway Basis Reference(s):
: 1. ECA-1.2 LOCA Outside Containment Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Page 237 of 56 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 237 of 245
: 2. E-1 Loss of Reactor or Secondary Coolant
: 3. NEI 99-01 CMT Integrity or Bypass Containment Loss 
 
[Document No.]
 
[Document No.]InsideRCP Seal CoolingInterface leakage Damper[Document No.]
Inside Reactor Building RCP Seal Cooling Interface leakage Damper Open valveOpen valveOpen valveFission Product Barrier Loss/Potential Loss Matrix and BasesFigure Reactor Interface leakage valve valve valve Fission Product Barrier Loss/Potential Loss Matrix and BasesFigure 1: ContainmentAuxiliary BuildingPenetrationOpenOpenATTACHMENT 2Page 36 of 56 Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Integrity or Bypass ExamplesAuxiliary Building OpenPenetration Damper Process MonitorOpen valve Open valve  ATTACHMENT 2 Page 36 of 56 Fission Product Barrier Loss/Potential Loss Matrix and BasesRev. 6 Integrity or Bypass ExamplesFilteOpen valve  Process Monitor Effluent Monitor Pump Fission Product Barrier Loss/Potential Loss Matrix and BasesIntegrity or Bypass ExamplesVent Closed Cooling WaterPump Fission Product Barrier Loss/Potential Loss Matrix and Bases Integrity or Bypass Examples Closed Cooling Water Airborne Monitor Area Monitor ~ !Page ~ !Page 238 of 245
 
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 239 of 245 Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: 1. CSFST Containment-RED path conditions met Definition(s):
None Basis: Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 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 Accident (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 potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the 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 barrier. Callaway Basis Reference(s):
: 1. CSF-1 Critical Safety Function Status Trees Containment Figure 5 2. Calc 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 Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 240 of 245 Barrier: Containment Category: D. CMT 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 of construction and radiolysis of aqueous solution in the core and sump. (ref. 1). 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. 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 occurred. 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 AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and ERFIS (ref. 3). The hydrogen monitors require a 2 hour warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 4, 5). If an actual 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 minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.
Callaway Basis Reference(s):
: 1. UFSAR Section 6.2 Containment Systems Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 241 of 245
: 2. FR-Z.4 Response to High Containment Hydrogen Concentration 3. FSAR Table 7A-3 (Sheet 32 Data Sheet 6.4)
: 4. OTN-GS-00001 Containment Hydrogen Control System
: 5. Calc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102 6. CA-3 Hydrogen Flammability in Containment
: 7. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B 
 
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 242 of 245 Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: 3. Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for  15 min. (Note 1, 9) Note 1: The Emergency Director 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. Definition(s):
None Basis: 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 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 fans are designed to start automatically in 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 full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 243 of 245 condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
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. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C 
 
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 244 of 245 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 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 "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. Callaway Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Loss 6.A  Plant-Specific&#xb6;Generic&#xb6; ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases [Document No.] Rev. 6 Page 245 of 245 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 Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
* Imminent barrier degradation exists if the degradation will likely occur within 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. Callaway Basis Reference(s):
: 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A
 
Plant-Specific&#xb6;Generic&#xb6; SU4.2 SU3 2
 
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